ML082470623

License Amendment Request: Appendix K Measurement Uncertainty Recapture - Power Uprate Request

ML082470623
Person / Time
Site:	Calvert Cliffs
Issue date:	08/29/2008
From:	Bauder D Constellation Energy Group, Nuclear Generation Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RIS-02-003
Download: ML082470623 (132)

Text

Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 0Constellation Energy-Nuclear Generation Group August 29, 2008 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos. I & 2; Docket Nos. 50-317 & 50-318 License Amendment Request: Appendix K Measurement Uncertainty Recapture-Power Uprate Request

REFERENCES:

(a) Nuclear Regulatory Commission Regulatory Issue Summary (RIS)2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," dated January 31, 2002 (b) SECY-04-0104, Status Report on Power Uprates Pursuant to 10 CFR 50.90, the Calvert Cliffs Nuclear Power Plant, Inc. hereby requests an amendment to the Renewed Operating License Nos. DPR-53 and DPR-69 to increase the licensed core power. Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 are currently licensed for a Rated Thermal Power of 2700 MWt. Based on the implementation of more accurate feedwater flow measurement instrumentation, approval is sought to increase the core power by 1.38 percent to 2737 MWt.The approach used in this amendment request follows that outlined in Reference (a). Reference (a)provides guidance on the scope and detail of the information that should be provided to the Nuclear Regulatory Commission for the review of measurement uncertainty recapture power uprate applications.

The significant hazards discussion and the technical basis for this proposed change are provided in Attachment (1). Attachment (2) provides the information delineated in Reference (a). Marked up pages of the affected Operating Licenses and Technical Specifications are provided in Attachment (3). The Technical Specification Bases will not require, any changes to be made.Based on expected Nuclear Regulatory Commission review timeframes as expressed in Reference (b), we request approval of this proposed change by March 1, 2009. Although this requested approval date does not impact continued operation of the Units at our current allowed power level (2700 MWt) it is needed to allow implementation of the proposed amendment following Unit 2 expected return to operation date following its 2009 refueling outage. We also request a 180 day implementation period for the approved amendment to allow sufficient time to implement procedure changes and operator training associated with this change.

Document Control Desk August 29, 2008 Page 2 Should you have questions regarding this matter, please contact Mr. Jay S. Gaines at (410) 495-5219.STATE OF MARYLAND COUNTY OF CALVERT: TO WIT: I, Douglas R. Bauder, being duly sworn, state that I am Plant General Manager -Calvert Cliffs Nuclear Power Plant, Inc. (CCNPP), and that I am duly authorized to execute and file this License Amendment Request on behalf of CCNPP. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements pare n.tbased on my personal knowledge, they are based upon information provided by other CNP mp11) ees and/or consultants.

Such information has been reviewed in accordance with compan prac and elieve it to be reliable.Subscribed and sworn before me, a Notary Public in and for the State of Maryland and County of C&I VeS ..., this 3 day of dtuaLLa 2008.WITNESS my Hand and Notarial Seal: My Commission Expires: N244/'NOTARY PUS6UC Q0" CwMY, Mwyd Up Cmm"6100upre01/01/10 Date DRB/KLG/bjd Attachments:

(1) Technical Basis and No Significant Hazards Consideration (2) Summary of Calvert Cliffs Nuclear Power Plant Measurement Uncertainty Recapture Evaluation Enclosure (1) CA006945, Revision 0000, Calorimetric Uncertainty Using the LEFM CheckPlus Flow Measurement System (3) Marked up Technical Specification Pages cc: D. V. Pickett, NRC S. J. Collins, NRC Resident Inspector, NRC S. Gray, DNR ATTACHMENT (1)TECHNICAL BASIS AND NO SIGNIFICANT HAZARDS CONSIDERATION TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions 5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

Calvert Cliffs Nuclear Power, Inc.August 29, 2008 ATTACHMENT (1)TECHNICAL BASIS AND SIGNIFICANT HAZARDS CONSIDERATION 1.0

SUMMARY

DESCRIPTION This letter requests an amendment to Renewed Operating License DPR-53 and DPR-69 for Calvert Cliffs Nuclear Power Plant (Calvert Cliffs) Unit Nos. 1 and 2, including Appendix A, Technical Specifications, to increase the licensed core power. Calvert Cliffs Units 1 and 2 are currently licensed for a Rated Thermal Power (RTP) of 2700 MWt. Through the use of more accurate feedwater flow measurement equipment, approval is sought to increase this core power by 1.38 percent to 2737 MWt. The power uprate is based on the use of the Caldon Leading Edge Flow Measurement (LEFM) CheckPlus system for determination of main feedwater flow and the associated determination of reactor power through the performance of the power calorimetric calculation currently required by Calvert Cliffs Technical Specifications.

2.0 DETAILED DESCRIPTION This proposed license amendment would revise the Calvert Cliffs Nuclear Power Plant Operating Licenses and Technical Specifications to increase the licensed power level to 2737 MWt, or 1.38 percent greater than the current level of 2700 MWt. The proposed changes, which are indicated on the marked up pages in Attachment (3), are described below: 1. Paragraph 2.C.(1) in Renewed Operating License Nos. DPR-53 and DPR-69 is revised to authorize operation at a steady-state reactor core power level not in excess of 2737 megawatts-thermal (100 percent power).2. The definition of RATED THERMAL POWER (RTP) in Technical Specification 1.1 is revised to reflect the increase from 2700 MWt to 2737 MWt.Calvert Cliffs Units 1 and 2 are presently licensed for an RTP of 2700 MWt. Through the use of more accurate feedwater flow measurement equipment, approval is sought to increase this core power by 1.38 percent to 2737 MWt.The approach used in this amendment request follows that outlined in Reference (1), Reference (1)provides guidance on the scope and detail of the information that should be provided to the Nuclear Regulatory Commission (NRC) for the review of measurement uncertainty recapture (MUR) power uprate applications.

The 1.38 percent core power MUR uprate for Calvert Cliffs is based on eliminating unnecessary analytical margin originally required for Emergency Core Cooling System (ECCS) evaluation models performed in accordance with the requirements set forth in Reference (2). The NRC has approved a change to the requirements of Reference (2) as described in the Federal Register (65 FR 34913, June 1, 2000). The change provides licensees with the option of maintaining the two percent power margin between the licensed power level and the assumed power level for the ECCS evaluation, or applying a reduced margin for ECCS evaluation.

For the reduced margin for ECCS evaluation case, the proposed alternative reduced margin must account for uncertainties due to power level instrumentation error.Based on the proposed use of the Caldon LEFM CheckPlus instrumentation with a power measurement uncertainty of less than 0.6 percent, it is proposed to reduce the licensed power uncertainty required by Reference (2). This results in the proposed increase of 1.38 percent (2737 MWt) in the Calvert Cliffs licensed power level using current NRC-approved methodologies.

The Caldon LEFM CheckPlus instrumentation provides a more accurate indication of feedwater flow (and correspondingly reactor thermal power) than assumed during the original development of Reference (2) requirements.

The improved thermal power measurement accuracy eliminates the need for the full two percent power margin assumed in Reference (2), thereby increasing the thermal power available for electrical generation.

I ATTACHMENT (1)TECHNICAL BASIS AND SIGNIFICANT HAZARDS CONSIDERATION 3.0 TECHNICAL EVALUATION The impact of the proposed power uprate on applicable systems, components, and safety analyses has been evaluated.

Attachment (2) summarizes the results of the comprehensive engineering review performed to evaluate the increase in the licensed core power from 2700 MWt to 2737 MWt. The results of this evaluation are provided in a format consistent with regulatory guidance provided in Reference (1).As discussed in Attachment (2), the evaluations and analyses have been completed to support an increase in RTP from 2700 MWt to 2737 MWt. In many cases an RTP of 2746 MWt (or a target power uprate of 1.7 percent) was used in order to -provide bounding input for these evaluations.

Currently, with the RTP of 2700 MWt, an analytical power level of 2754 MWt (102 percent of 2700 MWt) is used in the safety analysis.

With a requested revised RTP of 2737 MWt and a revised uncertainty, the analytical power level is unchanged at 2754 MWt.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The requirements for the ECCS Evaluation Models are set forth in 10 CFR Part 50, Appendix K. The NRC-approved a change to these requirements (Federal Register Notice 65 FR 34913, June 1, 2000) that provides licensee with the option of maintaining the two percent power measurement uncertainty in the ECCS analyses or of using a lower value provided the proposed alternative value has been demonstrated to account for the uncertainties due to power level instrumentation error.Calvert Cliffs proposes to increase RTP from 2700 MWt to 2737 MWt on both Units I and 2. The proposed change justifies use of an alternate power level other than 102% of RTP in the ECCS analysis based on the installation of a high accuracy feedwater flow instrumentation system (Caldon LEFM CheckPlus system) which results in a reduction of uncertainty in the power level measurement.

This resultant increase in RTP level is referred to as MUR. The analysis and detailed review conducted to support this requested power increase conforms to the guidance specified in NRCs Regulatory Issue Summary 2002-03.4.2 Precedent Below is a list of other facilities that have been granted approval for MUR power uprates involving the use of Caldon LEFM CheckPlus feedwater flow instrumentation includes: Facilit1y Amendment

1. (s) Approval Date Crystal River, Unit 3 228 December 26, 2007 Vogtle Electric Generating Plant, Units I & 2 149/129 February 27,2008 Cooper Nuclear Station 231 June 30, 2008 Davis Besse Nuclear Power Station, Unit 1 278 June 30, 2008 4.3 Significant Hazards Consideration Calvert Cliffs is proposing an amendment to the Facility Operating License and Technical Specifications that will increase the licensed power level from 2700 MWt to 2737 MWt based on the use of more accurate feedwater flow measurement equipment.

Use of the Caldon LEFM CheckPlus feedwater flow instrumentation reduces measurement uncertainty in the measurement system for determination of main feedwater flow and the associated determination of reactor power through the performance of the power calorimetric calculation currently required by Calvert Cliffs Technical Specifications.

The proposed changes have been evaluated against the standards in 10 CFR 50.92 and have been determined to not 2 ATTACHMENT (1)TECHNICAL BASIS AND SIGNIFICANT HAZARDS CONSIDERATION involve a significant hazards consideration in the operation of the facility in accordance with the proposed amendment.

1. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated In support of this measurement uncertainty recapture (MUR) power uprate, a comprehensive evaluation was performed for Nuclear Steam Supply System (NSSS), balance of plant systems and components, and analyses that could be affected by this change. A power calorimetric uncertainty calculation was performed, and the impact of increasing plant power by 1.38 percent on the plant's design and licensing basis was evaluated.

The result of these evaluations is that structures, systems, and components required to mitigate transients will continue to be capable of performing their design function at an uprated core power of 2737 MWt. In addition, an evaluation of the accident analyses demonstrates that applicable analysis acceptance criteria continue to be met. No accident initiators are affected by this uprate and no challenges to any plant safety barriers are created by this change.Therefore, operation of the facility in accordance with the proposed change will not involve a significant increase in the probability of an accident previously evaluated.

The proposed change does not affect the radiological release paths, the frequency of release, or the source-term for release for any accidents previously evaluated in the Updated Final Safety Analysis Report. Structures, systems, and components required to mitigate transients remain' capable of performing their design functions, and thus were found acceptable.

The reduced uncertainty in the feedwater flow input to the power calorimetric measurement ensures that applicable accident analyses acceptance criteria continue to be met in support of operation at a core power of 2737 MWt.Analyses performed to assess the effects of mass and energy remain valid. The source-terms used to assess radiological consequences have been reviewed and determined to bound operation at the uprated condition.

Therefore, operation of the facility in accordance with the proposed change will not involve a significant increase in the consequences of an accident previously evaluated.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

No new accident scenarios, failure mechanisms, or single-failures are introduced as a result of the proposed changes. The installation of the Caldon LEFM CheckPlus feedwater flow instrumentation system has been analyzed, and failures of this system will have no adverse effect on any safety-related system or any structures, systems, and components required for transient mitigation.

All structures, systems and components previously required for the mitigation of a transient remain capable of fulfilling their intended design functions.

The proposed changes have no adverse effects on any safety-related system or component and do not challenge the performance or integrity of any safety-related system.This change does not adversely affect any current system interfaces or create any new interfaces that could result in an accident or malfunction of a different kind than was previously evaluated.

Operating at a core power level of 2737 MWt does not create any new accident initiators or precursors.

The reduced uncertainty in the feedwater flow input to the power calorimetric measurement ensures that applicable accident analyses acceptance criteria continue to be met to support operation at a core power of 2737 MWt. Credible malfunctions continue to be bounded by 3 ATTACHMENT (1)TECHNICAL BASIS AND SIGNIFICANT HAZARDS CONSIDERATION the current accident analysis of record or evaluations that demonstrate that applicable acceptance criteria continue to be met.Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.The margins of safety associated with the MUR power uprate are those pertaining to core power.This includes those associated with the fuel cladding, Reactor Coolant System pressure boundary, and containment barriers.

A comprehensive engineering review was performed to evaluate the 1.38 percent increase in the licensed core power from 2700 MWt to 2737 MWt. The 1.38 percent increase required that revised NSSS design thermal and hydraulic parameters be established, which then served as the basis for all of the NSSS analyses and evaluations.

This engineering review concluded that no design modifications are required to accommodate the revised NSSS design conditions.

The NSSS components were evaluated and it was concluded that the NSSS components have sufficient margin to accommodate the 1.38 percent power uprate. The NSSS accident analyses were evaluated for the 1.38 percent power uprate. In all cases, the evaluations demonstrate that the applicable analyses acceptance criteria continue to be met. As a result, the margins of safety continue to be bounded by the current analyses of record for this change.Therefore, the proposed change does not involve a significant reduction in the margin of safety.Based on the above, Calvert Cliffs concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will hot be inimical to the common defense and security or to the health and safety of the public.5.0 ENVIRONMENTAL CONSIDERATION Implementation of the MUR power uprate is expected to result in a correspondingly small increase (no more than 1.38%) in general radiation levels and in the liquid and gaseous effluent releases.

This small increase will have minimal impact as existing site processes and practices are adequate to maintain offsite release concentrations and individual doses within the limits of 10 CFR Part 20 and 10 CFR Part 50, Appendix I.Calvert Cliffs, has determined that operation with the proposed amendment would not result in any significant change in the types, or significant increases in the amounts, of any effluents that may be released offsite, nor would it result in any significant increase in individual or cumulative occupational radiation exposure.

Therefore, the proposed amendment is eligible for categorical exclusion as set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed amendment.

4 ATTACHMENT (1)TECHNICAL BASIS AND SIGNIFICANT HAZARDS CONSIDERATION

6.0 REFERENCES

(1) Nuclear Regulatory Commission Regulatory Issue Summary 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," dated January 31, 2002 (2) 10 CFR Part 50, Appendix K, ECCS Evaluation Models 5 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION Calvert Cliffs Nuclear Power Plant, Inc.August 29, 2008 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION TABLE OF CONTENTS Page Section I Section II Section III Section IV Section V Section VI Section VII Section VIII Enclosure (1)Table of Contents List of Acronyms Introduction Feedwater Flow Measurement Technique and Power Measurement Uncertainty Accidents and Transients for which the Existing Analyses of Record Bound Plant Operation at the Proposed Increased Power Level Accidents and Transients for which the Existing Analyses of Record Do Not Bound Plant Operation at the Proposed Increased Power Level Mechanical/Structural/Material Component Integrity and Design Electrical Equipment Design System Design Other Changes to Technical Specifications, Protection System Settings,, and Emergency System Settings CA06945, Revision 0000, Calorimetric Uncertainty Using the LEFM CheckPlus Flow Measurement System ii v 1 13 34 36 64 67 77 80 i ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION ABB ABB-NV ABB-TV AC ADV AFAS AFW ALARA ANSI AOO AOP AOR AOV ASME AST ATWS BLPB BOP Calvert Cliffs CCW CE CEA CEDM CEOG CFR COLR CUF CVCS DAS DBA DBE DC DNB DNBR'D/P ECCS EM EOP EQ gpm HELB HFP HVAC HZP ICI IFBA LIST OF ACRONYMS Asea Brown Boveri, Inc.Asea Brown Boveri, Inc.-Non-Turbo Vane Asea Brown Boveri, Inc.-Turbo Vane Alternating Current Atmospheric Dump Valves Auxiliary Feedwater Actuation System Auxiliary Feedwater As Low As Reasonably Achievable American National Standards Institute Anticipated Operational Occurrence Abnormal Operating Procedures Analysis of Record / Analyses of Record Air-Operated Valve American Society of Mechanical Engineers Alternative Radiological Source Term Anticipated Transients Without SCRAM Branch Line Pipe Break Balance of Plant Calvert Cliffs Nuclear Power Plant, Inc.Component Cooling Water Combustion Engineering Control Element Assembly Control Element Drive Mechanism Combustion Engineering Owners Group Code of Federal Regulations Core Operating Limits Report Cumulative Usage Factor Chemical and Volume Control System Data Acquisition System Design Basis Accident Design Basis Event Direct Current Departure from Nucleate Boiling Departure from Nucleate Boiling Ratio Differential Pressure Emergency Core Cooling System Evaluation Model Emergency Operating Procedures Environmental Qualification gallons per minute High Energy Line Break Hot Full Power Heating, Ventilation, and Air Conditioning Hot Zero Power Incore Instrumentation Integral Fuel Burnable Absorber ii ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION LIST OF ACRONYMS ISI Inservice Inspection IST Inservice Testing Ke plastic strain correction factor LBB Leak Before Break LBLOCA Large Break Loss-of-Coolant Accident.LCO Limiting Condition for Operation LEFM Leading Edge Flow Measurement LHR Linear Heat Rate LOCA Loss-of-Coolant Accident LOSP Loss of Secondary Pressure LPSI Low Pressure Safety Injection MCLB Main Coolant Loop Break MNSA Mechanical Nozzle Seal Assembly MOV Motor-Operated Valve MSIV Main Steam Isolation Valve MSLB Main Steam Line Break MSS Main Steam System MSSV Main Steam Safety Valve MTC Moderator Temperature Coefficient MUR Measurement Uncertainty Recapture MVA MegaVolt Ampere MVAR MegaVolt Ampere Reactive MWt Megawatt Thermal NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System OBE Operating Basis Earthquake ODCM Offsite Dose Calculation Manual OOS Out-of-Service PLHGR Peak Linear heat Generation Rate PWR Pressurized Water Reactor RCP Reactor Coolant Pump RCS Reactor Coolant System RIS Regulatory Issue Summary RPS Reactor Protective System RTD Resistance Temperature Detector RTP Rated Thermal Power RV Reactor Vessel RVI Reactor Vessel Internal S2M Supplement 2 to CENPD-137 Evaluation Model SAFDL Specified Acceptable Fuel Design Limits SBLOCA Small Break Loss-of-Coolant Accident SDC Shutdown Cooling SER -Safety Evaluation Report SFPC Spent Fuel Pool Cooling SG Steam Generator SI Safety Injection iii ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION LIST OF ACRONYMS SIT Safety Injection Tank SLB Steam Line Break Sm Primary Membrane Stress SRW Service Water SW Saltwater Tave Vessel Average Coolant Temperature Tcold Vessel/Core/Inlet Temperature Thor Vessel Outlet Temperature TM/LP Thermal Margin/Low Pressure TRM Technical Requirements Manual UF Usage Factor UFM Ultrasonic Flow Measurement UFSAR Updated Final Safety Analysis Report VAP Value Added Fuel ZrB 2 Zirconium Diboride iv ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION INTRODIUCTLON BACKGROUND AND REASON FOR PROPOSED CHANGE Calvert Cliffs Nuclear Power Plant (Calvert Cliffs) Units 1 and 2 are presently licensed for a Rated Thermal Power (RTP) of 2700 MWt. Through the use of more accurate feedwater flow measurement equipment, approval is sought to increase this core power by 1.38% to 2737 MWt. The impact of a 1.38% core power uprate for applicable systems, components, and safety analyses has been evaluated.

The analyses and evaluations were performed for both Calvert Cliffs Units I and 2. In some cases where cycle specific data is needed, the analyses/evaluations targeted Unit I as the lead unit for the Measurement Uncertainty Recapture (MUR) power uprate. However because of the timing involved, Unit 2 will likely be the first unit to implement the MUR power uprate. Confirmation of the applicability of the cycle specific analyses and evaluations on Unit 2 for this operating cycle, and all subsequent cycles of Units I and 2, are performed as part of the normal reload design process.The approach used in this amendment request follows that outlined in Nuclear Regulatory Commission (NRC) Regulatory Issue Summary (RIS) 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications" dated January 31, 2002. Regulatory Issue Summary 2002-03 provided guidance on the scope and detail of the information that should be provided to the NRC for the review of MUR power uprate applications.

The 1.38% MUR power uprate for Calvert Cliffs is based on eliminating unnecessary analytical margin originally required for Emergency Core Cooling System (ECCS) evaluation models (EMs) performed in accordance with the requirements set forth in the Code of Federal Regulations (CFR), 10 CFR Part 50, Appendix K (ECCS).As discussed in detail in Section II, the evaluations and analyses described have been completed to support an increase in RTP from 2700 MWt to 2737 MWt. In many cases an RTP of 2746 MWt (or a target power uprate of 1.7%) was used in order to provide bounding input for these evaluations.

Currently, with the RTP of 2700 MWt, the analytical power level of 2754 MWt (102% of 2700 MWt) is used in the safety analysis.

With a revised RTP of 2737 MWt and a revised uncertainty, the analytical power level is unchanged at 2754 MWt.The NRC has approved a change to the requirements of 10 CFR Part 50, Appendix K [65 FR 34913, June 1, 2000]. The change provides licensees with the option of maintaining the two-percent power margin between the licensed power level and the assumed power level for the ECCS evaluation, or applying a reduced margin for ECCS evaluation.

For the reduced margin for ECCS evaluation case, the proposed alternative reduced margin must account for uncertainties due to power level instrumentation error. Based on the proposed use of the Caldon Leading Edge Flow Measurement (LEFM) CheckPlus instrumentation with a power measurement uncertainty of less than 0.6%, it is proposed to reduce the licensed power uncertainty required by 10 CFR Part 50, Appendix K. This results in the proposed increase of 1.38% in the Calvert Cliffs licensed power level using current NRC approved methodologies.

The Caldon LEFM CheckPlus instrumentation provides a more accurate indication of feedwater flow (and correspondingly reactor thermal power) than assumed during the development of Appendix K requirements.

Technical support for this conclusion is discussed in detail in the Caldon LEFM CheckPlus Topical Reports (References I-1 and 1-2). The improved thermal power measurement accuracy eliminates the need for the full 2% power margin assumed in Appendix K, thereby increasing the thermal power available for electrical generation.

v ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION The desired power increase of 1.38% will be accomplished by increasing the electrical demand on the turbine-generator.

As a result of this demand increase, steam flow will increase and the resultant steam pressure will decrease.

The Reactor Coolant System (RCS) nominal cold leg temperature will remain constant while ,the hot leg temperature will increase slightly in response to the increased steam flow demand. As a result, the RCS average temperature will increase slightly.Procedures for maintenance and calibration of the Caldon LEFM CheckPlus system will be enhanced per the design control process based on the vendor's recommendations.

Should the Caldon LEFM CheckPlus system be unavailable, the main steam or feedwater flow venturis can be used to measure flow rate in the feedwater system, as is done currently.

If the Caldon LEFM CheckPlus system is not functional, the power limit will be administratively controlled at a level consistent with the accuracy of the available instrumentation as described in this amendment request. The power limit reduction requirement for the Caldon LEFM CheckPlus system out-of-service (OOS) will be incorporated into the Calvert Cliffs Technical Requirements Manual (TRM).DESCRIPTION OF PROPOSED CHANGE The proposed license amendment would revise the Calvert Cliffs Operating Licenses and Technical Specifications to reflect an increase in core power level by 1.38% to 2737 MWt. The power uprate is based on the use of the Caldon LEFM CheckPlus system for determination of main feedwater flow and the associated determination of reactor power through the performance of the power calorimetric calculation currently required by Calvert Cliffs Technical Specifications.

The proposed changes are identified on the markups of the current Calvert Cliffs Operating Licenses and Technical Specification pages.Calvert Cliffs notes that various Combustion Engineering (CE) topical reports that are part of the Calvert Cliffs licensing basis (Technical Specification 5.6.5), consistent with 10 CFR Part 50, Appendix K may have included explicit references to their use of "102% of licensed core power levels." These topical reports describe the NRC-approved methodologies which support the Calvert Cliffs safety analyses, including the small break and large break loss-of-coolant accident (LOCA) analyses.

Along with the proposal to increase the reactor thermal power to 2737 MWt, Calvert Cliffs requests continued use of these topical reports. Calvert Cliffs does not consider that these topical reports require revision to reflect this requested power uprate. Rather, it will be understood that those statements refer to the Appendix K margin and the original licensed power level. Calvert Cliffs proposes that these topical reports be approved for use consistent with this license amendment request, and further, the NRC acknowledges that the change in the power uncertainty does not constitute a significant change, as defined in 10 CFR 50.46 and 10 CFR Part 50, Appendix K, to these topical reports.GENERAL LICENSING APPROACH FOR PLANT ANALYSIS USING PLANT POWER LEVEL The MUR power uprate program for Calvert Cliffs as described addresses Nuclear Steam Supply System (NSSS) performance parameters, design transients, systems, components, accidents, and nuclear fuel as well as interfaces between the NSSS and Balance of Plant (BOP) systems. No new analytical techniques have been used to support the MUR power uprate project. The key points include the use of:* Well-defined analysis input assumptions/parameter values* Currently approved analytical techniques

• Applicable licensing criteria and standards vi ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION The evaluations and analyses described have been completed in order to bound an increase in RTP from 2700 MWt to 2737 MWt, or a 1.38% increase.

The RTP used for many evaluations targeted a bounding power uprate of 1.7% with MUR power uprate, or 2746 MWt. Currently, with the RTP of 2700 MWt, the analytical power level of 2754 MWt (102% of 2700 MWt) is used in the safety analysis.

With a revised RTP of 2737 MWt and a revised uncertainty, the analytical power level is unchanged at 2754 MWt.Section I provides a description of the feedwater flow measurement system that will be installed on both units.Section I also provides a summary of the overall thermal power measurement uncertainty.

Section II provides the results of the accident analyses and evaluations performed for the LOCA and non-LOCA transients.

Section II also summarizes the containment accident analyses and evaluations and the radiological consequence evaluations.

Section III provides results for accidents and transients for which the existing analyses of record (AOR)do not bound plant operation at the proposed uprated power level.Section IV of.this report discusses the revised NSSS design thermal and hydraulic parameters that were modified as a result of the MUR power uprate and that serve as the basis for all of the NSSS analyses and evaluations.

In addition this section discusses the effect of the power uprate on the structural integrity of major plant components.

Section IV also contains the results of the fuel-related analyses.Section V provides an analysis of the effects of the power uprate on the Calvert Cliffs electrical power systems.Section VI presents information on the impact of the proposed power uprate on the system design [e.g., safety injection (SI), shutdown cooling (SDC), and control systems] and components

[e.g., reactor vessel (RV), pressurizer, Reactor Coolant Pumps (RCPs), steam generator (SG), and NSSS auxiliary equipment]

and the evaluations completed for the revised design conditions.

Section VI also summarizes the effects of the uprate on the BOP (secondary) systems based upon a heat balance evaluation.

Section VII evaluates the impact of the power uprate on various other areas including the impact on plant operations, the impact on the environment and the impact on occupational radiation exposure.Section VIII presents information on changes to Technical Specifications, protection system settings, and emergency system settings as a result of the proposed power uprate.The results of all of the analyses and evaluations performed demonstrate that all acceptance criteria continue to be met.vii ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION I. FEEDWATER FLOW MEASUREMENT TECHNIQUE AND POWER MEASUREMENT UNCERTAINTY 1.1 APPROVED TOPICAL REPORTS ON FEEDWATER FLOW MEASUREMENT TECHNIQUE The reference Topical Reports associated with the Caldon LEFM CheckPlus feedwater flow measurement system are as follows: 1. ER-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level using the LEFM Check System," dated March 1997 (Reference I- 1)2. ER-157P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check or CheckPlus System" dated October 2001 (Reference 1-2)1.2 NRC APPROVAL OF FEEDWATER FLOW MEASUREMENT TECHNIQUE The NRC approved the subject Topical Reports listed in Section 1.1 above on the following dates: 1. Reference 1-1, NRC Safety Evaluation Report (SER) dated March 8, 1999 2. Reference 1-2, NRC SER dated December 20, 2001 1.3 CALDON LEFM CHECKPLUS SYSTEM The feedwater flow measurement system to be installed at Calvert Cliffs is the Caldon LEFM CheckPlus ultrasonic multi-path transit time flow meter. The installation of this system will conform to the requirements of References 1-1 and 1-2. Subsequent reviews by the NRC, in Reference 1-3 found that the performance of the CheckPlus system was consistent with the topical reports, with one exception to further. evaluate the effects of transducer replacement on system uncertainty.

The exception was subsequently addressed'in Reference 1-4 and disseminated to the industry via Reference 1-5. The installation at Calvert Cliffs will include the additional uncertainty associated with transducer replacement, described in References 1-4 and I-5.The Caldon LEFM CheckPlus System to be installed at Calvert Cliffs consists of two spool piece measurement sections per unit with one spool piece installed in the 16" feedwater header for each SG.Each spool piece consists of 16 transducers, arranged in two planes with four pairs of transducers in each plane. The transducers are located in wells, such that a transducer may be removed at power without disturbing the pressure boundary of the spool piece. The installation locations conform to the requirements of References 1-1 and 1-2. The measurement sections will be installed in accordance with approved Calvert Cliffs procedures and work controls processes to achieve installation tolerances within the bounds stated in the Caldon uncertainty analysis.A cabinet mounted Caldon LEFM CheckPlus electronic unit will also be installed in the Turbine Building, in the vicinity of the measurement sections.

One cabinet will be installed per unit. The cabinets contain an integral air conditioning unit to maintain an acceptable operating environment.

Two pressure transmitters meeting the uncertainty requirements of the Caldon Topical Reports will be installed in each feedwater header in the vicinity of the spool pieces. The pressure transmitters provide input of feedwater pressure to the electronic unit for the calculation of feedwater flow.The Caldon LEFM CheckPlus Systems determine feedwater parameters for feedwater mass flow, feedwater temperature, and feedwater pressure to be used for the continuous calculation of secondary I ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION calorimetric power. The measured feedwater parameters are communicated to the Plant Computer and Data Acquisition System (DAS) over the Plant Data Network for use in the calorimetric power algorithm.

Each system incorporates self-verification features to ensure that the system continually operates within the design basis uncertainty analysis.

Diagnostic and signal quality data is communicated to the DAS to allow monitoring of degradation of the Caldon LEFM CheckPlus System. The system triggers control room annunciation via the Plant Computer when conditions are at a state which could impact the flow measurement uncertainty.

The LEFM CheckPlus measurement sections are calibrated in a site-specific model test at Alden Research Laboratories with all calibration standards traceable to National Institute of Standards and Technology standards.

The site-specific test plan provides meter factor calibration data over a wide range of hydraulic test conditions intended to envelope the expected hydraulic conditions at the installation locations.

The tests include plant piping modeling and parametric variations of those models, straight pipe testing, and inducement of extreme swirl conditions.

The meter factor data, determined by comparing the Alden Lab reference standard to the flow as measured by the Caldon LEFM CheckPlus System, is collected for each piping configuration at various flow rates. Measurement of the hydraulic profile, called the flatness ratio, is also collected at each flow rate. The meter factor versus flatness ratio is plotted for all conditions and all flow rates and compared to analytically derived expected performance curves for quality control purposes.

These data provide a quantitative measure of the Caldon LEFM CheckPlus Meter Factor verses the actual velocity profile encountered and determines the meter uncertainty to be used in the overall calorimetric uncertainty.

LEFM CheckPlus System Controls, Displays, and Alarms There are no LEFM CheckPlus System controls available in the Control Room. All control functions reside locally at the LEFM CheckPlus system cabinets located in the Turbine Building.Control Room operators can select the LEFM CheckPlus System output as the source of input data for the Plant Computer calculation of calorimetric calculation via a control room display interface.

The results of the calorimetric calculation are displayed on the Plant Computer to Control Room operators.

System alarms trigger an alarm resulting in control room annunciation.

There are no hardwired alarms from the LEFM CheckPlus System cabinet to the Control Room. The following conditions trigger the alarm:* LEFM CheckPlus System Meter Status Not Normal -the system and meter status (Normal, Alert, Failed) are communicated to the DAS. An Alert or Failed condition indicates a condition that may adversely affect the uncertainty of the LEFM CheckPlus System mass flow rate determination and triggers the Plant Computer alarm and control room annunciation.

Upon receipt of this alarm, the LEFM CheckPlus System is considered either in a degraded status or OOS.* Loss of Communication

-a communications failure from the LEFM CheckPlus System to the Plant Computer triggers the Plant Computer alarm and control room annunciation.

Upon receipt of this alarm, the LEFM CheckPlus System is considered OOS." LEFM CheckPlus System Cabinet High Temperature

-a cabinet high temperature condition also triggers the Plant Computer alarm and control room annunciation.

If the maximum temperature limit is exceeded, the LEFM CheckPlus System is considered OOS.Guidance will be provided to identify the actions to be taken by the Control Room staff upon alarm annunciation.

Detailed LEFM CheckPlus System process and diagnostic data, communicated to the 2 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION DAS, is also be available for use by operations staff for diagnosis of system alarms. The process and diagnostic data is also available locally at the LEFM CheckPlus cabinet.Refer to Sections 1.7 and 1.8 for additional information regarding operation in a degraded or OOS condition.

1.4 COMPLIANCE WITH NUCLEAR REGULATORY COMMISSION SAFETY EVALUATION REPORT The installation of the Caldon LEFM CheckPlus flow measurement system at Calvert Cliffs is consistent with References I-1 and 1-2. In addition to the installation requirements, the NRC identified in Reference 1-6, the following four criteria that must be addressed by licensees requesting a license amendment based on the Topical Reports. Calvert Cliffs meets the four criteria as described belbw.Criterion I Discuss maintenance and calibration procedures that will be implemented with the incorporation of the Caldon LEFM CheckPlus, including processes and contingencies for inoperable Caldon LEFM CheckPlus instrumentation, and the effect on thermal power measurements and plant operation.

Response to Criterion 1 Implementation of the power uprate license amendment includes developing the necessary procedures and documents required for operation, maintenance, calibration, testing, and training at the uprated power level with the new Caldon LEFM CheckPlus System. Plant procedures will be revised to incorporate the vendor's maintenance and calibration requirements prior to declaring the Caldon LEFM CheckPlus System operational and raising reactor core power above 2700 MWt (98.6% of proposed RTP). The incorporation of, and continued adherence to, these requirements assure that the Caldon LEFM CheckPlus System is properly maintained and calibrated.

Calibration and maintenance are further discussed in Section 1.6 below.Administrative and procedural controls will be established to provide guidance to operators in the event that Caldon LEFM CheckPlus system is unavailable.

Contingency plans for operation of the plant with the Caldon LEFM CheckPlus degraded or OOS are described in detail in Sections 1.7 and 1.8 below.Criterion 2 For a plant that currently has Caldon LEFM CheckPlus system installed, provide an evaluation of the operational and maintenance history of the installed instrumentation and confirmation that the installed instrumentation is representative of the LEFM system and bound the analysis and assumptions set forth in Reference I-1.Response to Criterion 2 The Caldon LEFM CheckPlus system is not currently installed at Calvert Cliffs.Criterion 3 Confirm that the methodology used to calculate the uncertainty of the Caldon LEFM CheckPlus system in comparison to current feedwater instrumentation is based on accepted plant setpoint methodology (with regard to the development of instrument uncertainty).

If an alternative approach is used, the application 3

ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION should be justified and applied to both venturi and ultrasonic flow measurement (UFM) instrumentation installations for comparison.

Response to Criterion 3 The methodology used to calculate the uncertainty of the Caldon LEFM CheckPlus system is consistent with the approved Topical Reports. The Topical Reports have been reviewed by site personnel and found to be consistent with Calvert Cliffs engineering standards, derived from Reference 1-7 and consistent with Reference 1-8. An alternative methodology is not used.Using site standards, uncertainties for parameters that are not statistically independent are arithmetically summed, then statistically combined with other parameters.

Random uncertainties are combined using the Square Root Sum of Squares approach.

Systematic biases are then added to the result to determine the overall uncertainty.

This methodology is consistent with the vendor determination of LEFM CheckPlus System uncertainty, described in the topical reports.Criterion 4 For plants where the ultrasonic meter (including Caldon LEFM CheckPlus) was not installed and flow elements calibrated to a site-specific piping configuration (flow profiles and meter factors not representative of the plant specific installation), additional justification should be provided for its use.The justification should show that the meter installation is either independent of the plant specific flow profile for the stated accuracy, or that the installation can be shown to be equivalent to known calibrations and plant configurations for the specific installation including the propagation of flow profile effects at higher Reynolds numbers. Additionally, for previously installed calibrated elements, confirm that the piping configuration remains bounding for the original UFM installation and calibration assumptions.

Response to Criterion 4 The Caldon LEFM CheckPlus System will be calibrated using a site-specific piping configuration at Alden Research Laboratories.

Testing will be witnessed by Calvert Cliffs personnel.

The site-specific test plan provides meter factor calibration data over a wide range of hydraulic test conditions intended to envelope the expected hydraulic conditions at the installation locations.

The results of the tests will be used as the basis for the vendor uncertainty reports and will be provided to Calvert Cliffs. The calibration meter factor and the uncertainty in the calibration factor are based upon these reports.Since the calibration of the Caldon LEFM CheckPlus measurement sections has not been completed, a flow measurement uncertainty of +/- 0.5% flow has been assumed to support the requested uprate.Furthermore, the calculation is based on using +/- 1.88'F uncertainty using the existing feedwater resistance temperature detectors (RTDs) for feedwater enthalpy and not the more precise temperature measurement available using the LEFM CheckPlus System. These assumptions are very conservative as the Caldon LEFM CheckPlus System is capable of a flow measurement uncertainty on the order of+/- 0.3% with a temperature measurement uncertainty of +/- 0.67F.Final acceptance of the Calvert Cliffs specific uncertainty analysis occurs after completion of the commissioning process. The commissioning process verifies that the in-situ test data is bounded by the calibration test data (see Appendix F of Reference I-1). This step provides final positive confirmation that actual performance is within the bounds established for the instrumentation.

Final commissioning of the Caldon LEFM CheckPlus Systems is expected to be completed following the 2009 Unit 2 spring refueling outage and the 2010 Unit 1 spring refueling outage.4 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION 1.5 THERMAL POWER MEASUREMENT UNCERTAINTY The impact of the LEFM CheckPlus system on the overall thermal power measurement uncertainty is presented in Enclosure (1). Since the calibration of the LEFM CheckPlus measurement sections has not been completed, conservative assumptions for flow and temperature measurement uncertainty (as detailed in the Response to Criterion 4 section above) are used in calculating the overall thermal power measurements.

These assumptions will be confirmed during acceptance of the vendor uncertainty reports by Calvert Cliffs.Upon receipt of the vendor calibration reports, the calorimetric uncertainty assessment will be revised, if necessary, to determine the available margin at the uprated power of 2737 MWt. The vendor's site-specific uncertainty analysis includes uncertainty associated with transducer replacement as required by Reference 1-3 and described in References 1-4 and 1-5.Tables I-1 and 1-2 summarize the core thermal power measurement uncertainty in percentage of the proposed uprated power of 2737 MWt for Calvert Cliffs for each input to the calorimetric calculation.

The parameter uncertainties in Table I-1 are based upon the instrumentation uncertainties listed in Table 1-2.For each random input in Table I-1, an effective contribution is also listed to permit the algebraic summing of the bias inputs with the random contribution to develop the combined uncertainty of each input. As shown in Table I-1, the sum of the effective contributions is equivalent to the square root of the sum of the squares of the random inputs.5 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION Table 1-1 Process Parameter Inputs to Secondary Calorimetric Calculation Random Inputs Effective Bias Inputs to Combined Combined Percent Random INPUT to Uncertainty, Contribution.

Uncertainty, Uncertainty, Uncertainty, % Contribution to MBTU/hr MBTU/hr MBTU/hr MBTU/hr RTP Uncertainty Feedwater Flow -29.1498 -25.3725 -25.3725 -0.2717% 61.679%Blowdown Flow -4.5779 -0.6258 .. -0.6258 -0.0067% 1.521%Feedwater Enthalpy:

15.2863 -6.9774-.

-6.9774 -0.0747% 16.962%Feedwater Temperature

-15.2682 -6.9610 -69610 -0.0745% 16:922%Feedwater Pressure -0.0884 -0.-0.0002 0.0000% 0.001%Plant Computer Calculation of Sub-cooled Liquid Enthalpy -0.73 70 -0.0162 -0.0162 -0.0002% 0.039%Steam Enthalpy:

4.1484 -0.5139 -1.1812 -1.6951 -0.0182% 4.121%Steam Pressure -4.0837 -0.4980 -1.1812 -1.6792 -0.0180% 4.082%Plant Computer Calculation of Saturated Vapor Enthalpy -0.7295 -0.0159 -0.0159 -0.0002% 0.039%Plant Computer Calculation of Saturated Liquid Enthalpy -0. 0091 0.0000 0.0000 0.0000% 0.000%Calorimetric Constants , -6.4657 -6.4657 -0.0692% 15.718%Totals -33.4895 (2) -33.4895 () -7.6469 3) -41.1365 (3) -0.4405% (3) 100.00%o Adjustments for miscellaneous heat addition and heat removal terms from the RCS, such as input from pressurizer heaters (2) Square Root Sum of Squares (3) Algebraic Sum 6 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION Table 1-2 Uncertainties of Inputs to Secondary Calorimetric Calculation Input Random Bias InputUncertainty Uncertainty Feedwater flow (assumed), % Flow +/-0.50%Feedwater pressure (assumed), psi +1- 15.00 Feedwater temperature (assumed), 'F +/- 1.88 Steam Pressure, psi +/- 19.80 + 3.40 Total Blowdown Flow, klbm/hr +/-8.1 Plant Computer Calculation of Enthalpies, BTU/lbm +/- 0.10 " " Calorimetric Constants, MBTU/hr (I) -6.465"7 ( Adjustments for miscellaneous heat addition and heat removal terms from the RCS, such as input from pressurizer heaters 1.6 CALIBRATION AND MAINTENANCE A. Maintaining Calibration:

Calibration and maintenance is performed by qualified Calvert Cliffs maintenance personnel using site procedures.

The site procedures will be enhanced using the Caldon LEFM CheckPlus technical manuals and work instructions.

All work is performed in accordance with site work control procedures.

Formal training on system operation and maintenance will be provided to the appropriate Calvert Cliffs personnel.

Operations training is conducted by qualified Calvert Cliffs personnel in accordance with approved site procedures for the performance of training.

All necessary training will be completed prior to commissioning of the Caldon LEFM CheckPlus System.Routine maintenance activities for the Caldon LEFM CheckPlus System include:* physical inspectionsof system components,* power supply checks, 0 analog input checks,* acoustic processor unit checks,* watchdog timer checks, 0 communications checks,* transducer cable checks,* dimensional checks, and* calibration of pressure transmitters for feedwater pressure input to the cabinet.Other instruments which provide input to the secondary calorimetric are periodically calibrated in accordance with approved site procedures to ensure reliable operation that satisfies the requirements of the calorimetric uncertainty calculation.

B. Controlling Software and Hardware Configuration:

The Caldon LEFM CheckPlus System is designed and manufactured in accordance with the vendor's 10 CFR Part 50, Appendix B, Quality Assurance Program and its Verification and Validation program. The vendor's Verification and Validation program satisfies the 7 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION requirements of References 1-9 and 1-10. In addition the program is consistent with guidance for software Verification and Validation in Reference I-i11.After installation, software and hardware configuration is controlled in accordance with site procedures and processes for software configuration control. Proposed changes to the software and hardware configuration for all components that provide input to the calorimetric calculation are evaluated in accordance with the approved engineering change process.C. Performing Corrective Actions: Reliability of the Caldon LEFM CheckPlus system and other calorimetric instrumentation is monitored by Calvert Cliffs Plant Engineering personnel.

Adverse performance trends, failed preventive maintenance, or other observed equipment deficiencies are documented and resolved in accordance with the site's corrective action process.Any needed corrective maintenance is performed by qualified Calvert Cliffs maintenance personnel.

D. Reporting Deficiencies to the Manufacturer:

Corrective action procedures include instructions for notification of deficiencies and error reporting.

Equipment manufacturers are contacted as required to correct the deficiency.

E. Receiving and Addressing Manufacturer Deficiency Reports: Manufacturer deficiency reports are reviewed and dispositioned in accordance with the site's corrective action program. In addition, incoming Institute of Nuclear Power Operations Operating Experience are reviewed by site personnel for applicability.

Those deficiencies applicable to Calvert Cliffs are documented under the site's corrective action process.1.7 OUTAGE TIME Each of the Caldon LEFM CheckPlus Systems to be installed will consist of two measurement sections.One measurement section is installed in the feedwater header to each SG. Each measurement section consists of two planes of transducers with four pairs of transducers in each plane, as described in Reference 1-2. The transducers provide input to the electronic unit cabinet, which consists of two subsystems of electronics hardware.

Each subsystem receives input from one plane of the measurement sections.

Outputs from the electronic unit are provided to the Plant Computer via the Plant Data Network and DAS for the calculation of calorimetric power. Programmed logic in the DAS and Plant Computer, alert operators when the system is in a degraded or OOS condition.

The following conditions trigger the Plant Computer alarm: LEFM CheckPlus System Meter Status Not Normal -the meter status (Normal, Alert, Failed) is communicated to the DAS and Plant Computer.

A meter status of other than normal triggers the Plant Computer alarm. The meter status is determined from a series of on-line self-diagnostics to verify that the system is operating within its design basis uncertainty limits. The following conditions result in a meter status of other than normal:-failure of one or more transducer paths,-velocity profile out of limits,-analog input out of limits,-system uncertainty out of limits.8 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION

• Loss of communication from the LEFM CheckPlus System to the Plant Computer.* Cabinet temperature exceeds limit.Guidance will be provided to identify the actions to be taken by the Control Room staff upon alarm annunciation.

If the system is degraded or OOS, time accrues against the allowable outage times. Upon reaching the limit for the allowable outage time, the maximum power limit will be reduced to the pre-uprate licensed power limit of 2700 MWt (98.6% proposed RTP). Power is adjusted, as required, to ensure the pre-uprate licensed power limit is not exceeded.Three outage times are proposed: If the LEFM CheckPlus System is in a degraded condition with the Plant Computer available to perform the secondary calorimetric calculation, the allowable outage time is 30 days.If the LEFM CheckPlus System is OOS with the Plant Computer available to perform the secondary calorimetric calculation, the allowable outage time is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, provided steady-state conditions exist. Steady-state conditions are defined as power variations of less than 10% from the initial power level when the system is declared OOS.* If the Plant Computer is unavailable or if another input to the secondary calorimetric calculation fails (other than the LEFM CheckPlus System), the allowable outage time is less than or equal to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.Allowable outage times will be described in the TRM. If the site-specific uncertainty analysis for the LEFM CheckPlus System does not support operation in a degraded condition, the 30--day outage time will not be adopted.LEFM CheckPlus System Degraded, Plant Computer Available A 30-day outage time is proposed if the LEFM CheckPlus System is degraded but the Plant Computer is available to perform the secondary calorimetric calculation.

The system is considered to be degraded when an alert condition is detected and reported by the system, resulting in control room annunciation.

The site-specific uncertainty calculation for the LEFM CheckPlus System includes uncertainty with the system in an alert condition.

If the resultant calorimetric uncertainty supports the proposed uprate, operation in the degraded condition can theoretically continue indefinitely, although at a reduced margin.However, operation in a degraded condition is limited to 30 days to ensure that the system is restored to a fully operational status. If an alert condition is detected, an operator verifies the cause of the alarm and determines if the system can continue to be operated in the degraded status.As described in Reference 1-2, the Caldon LEFM CheckPlus System consists of subsystems of electronic hardware.

An alert condition basically informs the operator of the malfunction of a single subsystem, resulting in a slight increase to calorimetric uncertainty.

In this condition, the system basically operates as the LEFM Check System described in References I-1 and 1-2, capable of supporting uprates on the order of the requested 1.38% uprate. However, if the site-specific uncertainty analysis for the LEFM CheckPlus System does not support the uprate, the 30-day outage time will not be adopted.If Calvert Cliffs is unable to restore the LEFM CheckPlus system to full operation within the 30-day outage window, operators take action as indicated in Section 1.8 below.9 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION LEFM CheckPlus System OOS, Plant Computer Available A 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> outage time is proposed if the LEFM CheckPlus System is OOS but the Plant Computer is available to perform the secondary calorimetric calculation.

The system is considered to be OOS when either a fail condition is detected and reported by the system failure or when communication with the system is lost, resulting in control room annunciation.

The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> outage time is based upon:* Calculation of calorimetric power using the. Plant Computer from alternate plant instrumentation.

The operator can select an alternate set of parameters in lieu of the output of the Caldon LEFM CheckPlus System to calculate calorimetric power for feedwater flow, temperature, and pressure.Existing plant instrumentation, such as the feedwater venturis, currently being used to calculate secondary calorimetric power, is used for the alternate set of parameters.

Normalizing the alternate input for feedwater flow and temperature to the Caldon LEFM CheckPlus feedwater flow and temperature.

A rolling average of the ratio of the LEFM CheckPlus input to the alternate input is calculated on the Plant Computer.

When the alternate set of parameters is selected, the last known good value of the average-ratios will be applied such that the output of the calorimetric calculation using the alternate parameters closely matches the output of the calculation using the Caldon LEFM CheckPlus System. As shown in Table I-1, the calorimetric calculation is not sensitive to changes in feedwater pressure, such that no correction is necessary to feedwater pressure.Unlikely occurrence of venturi nozzle fouling or defouling.

Calvert Cliffs does not have a history of venturi nozzle fouling and subsequent defouling.

Therefore, no change in calorimetric output from fouling or defouling is anticipated during the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OOS time. With the LEFM CheckPlus System OOS, alternate indications such as turbine first stage pressure and feedwater temperature, will be used to ensure that plant power is not adjusted to account for venturi nozzle defouling, in theunlikely event fouling exists. Adjustments based on nozzle fouling, should it occur, would result in a conservative adjustment to calorimetric power.Negligible instrument drift. Instrument drift over a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period is negligible and can be verified using alternate plant instrumentation such as turbine first stage pressure.Anticipated margin. The assumed values for feedwater flow uncertainty and feedwater temperature uncertainty to support the requested 1.38% uprate are more conservative than typical values for the LEFM CheckPlus System, which can be used to support uprates on the order of 1.6% to 1.7%.When the calorimetric uncertainty assessment is revised to incorporate the vendor calibration reports, the calorimetric uncertainty is reduced, increasing the available margin.Most repairs to the Caldon LEFM CheckPlus System are expected to be completed within a shift. The 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> gives plant personnel sufficient time to diagnose, plan, implement, .and verify repairs to the system. If repairs are not completed within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> window, operators take action as indicated in Section 1.8 below.Plant Computer Unavailable An outage time less than or equal to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is proposed if the Plant Computer is unavailable or if another input to the secondary calorimetric calculation fails, regardless of the status of the Caldon LEFM CheckPlus System. The outage time is based upon: The minimum frequency for the calibration of the power range nuclear instrumentation in accordance with Technical Specification Surveillance Requirement 3.3.1.2. Per Technical Specification Surveillance Requirement 3.3.1.2, the power range nuclear instruments are adjusted every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> based on the reactor thermal power calculation.

Therefore, the actual duration of 10 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION the allowable outage time is determined from the next required adjustment of the power range nuclear instruments after the failure is identified.

• The precision of the Plant Computer calculation is required to support the increased power level.Without the Plant Computer, the uncertainty of alternate indications that may be used to calculate calorimetric power exceeds the uncertainty required to support the power uprate. Additionally, averaging of the calorimetric calculation is no longer available." The failure of shared inputs to the calorimetric calculation.

Alternate inputs are available only for feedwater flow, temperature, and pressure.

Other inputs, such as steam pressure, do not have alternate inputs. If a shared input fails, calorimetric power cannot be calculated on the Plant Computer.Occasional bad quality data is expected and would not result in entrance into the OOS time unless the bad quality data resulted in bad quality for the four hour averaged calorimetric power calculation.

If Calvert Cliffs is unable to restore the Plant Computer to normal operation within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> window, operators take action as indicated in Section 1.8 below.1.8 OPERATOR ACTION TO REDUCE POWER For each of the three outage times indicated in Section 1-7, if necessary repairs are not completed within the allowed outage time window, operators take action to limit the maximum thermal power limit to the pre-uprate licensed power limit of 2700 MWt. One additional restraint on maximum power operation will be placed whenever a unit is within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> outage window due to the Caldon LEFM CheckPlus system being OOS. In this situation if the plant experiences a power change of more than 10% power, the maximum thermal power limit will be limited to the pre-uprate licensed power limit of 2700 MWt.Although power changes have not been shown as having a significant effect on the alternate calorimetric instrumentation, this conservative action ensures that a plant transient does not adversely impact the accuracy of the alternate calorimetric instrumentation.

Calvert Cliffs intends to document, within the site's TRM, necessary operator actions to address the instances when the Caldon LEFM CheckPlus System is not available to provide the feedwater flow element inputs to the heat balanced calorimetric algorithm power measurement, as well as actions to be taken if these inputs are not restored in the allowed time. Operator actions are captured in the TRM vice the Technical Specifications as the feedwater flow element inputs to the heat balance calorimetric algorithm do not meet the criteria of 10 CFR 50.36(d)(2)(ii) for establishing a Technical Specification Limiting Condition for Operation (LCO) as indicated below.Criterion I The Caldon LEFM CheckPlus feedwater flow element inputs are not used to detect and indicate abnormal degradation of the reactor coolant pressure boundary.Criterion 2 The Caldon LEFM CheckPlus feedwater flow element inputs are not initial conditions of a design basis accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.11 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION Criterion 3 The Caldon LEFM CheckPlus feedwater flow element inputs are not part of the primary success path and do not function or actuate to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.Criterion 4 In the event of the Caldon LEFM CheckPlus ultrasonic feedwater flow element inputs not being available for the heat balance calorimetric algorithm, the inputs will be determined by alternate instrumentation thus, the Caldon LEFM CheckPlus ultrasonic feedwater flow element inputs are not significant to public health and safety.It is therefore concluded that an LCO is not required to be included in the Technical Specifications in accordance with 10 CFR 50.36(d)(2)(ii) to address the functional requirements for the Caldon LEFM CheckPlus feedwater flow element inputs to the heat balance calorimetric algorithm.

1.9 REFERENCES

I-1 ER-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM Check System," dated March 1997 approved by NRC SER, dated March 8, 1999 1-2 ER-157P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check or CheckPlus System," Revision 5, dated October 2001, approved by NRC SER, dated December 20, 2001 1-3 Letter from B.E. Thomas (NRC) to Mr. E.M. Hauser (Caldon, Inc.), dated July 5, 2006,"Evaluation of the Hydraulic Aspects of the Caldon Leading Edge Flow Measurement (LEFM)Check and CheckPlus Ultrasonic Flow Meters (UFMs) (TAC No. MC6424)," Project No. 1311 1-4 ER-551P, "LEFM CheckPlus Transducer Installation Sensitivity," Revision 3, dated April 2008 I-5 Customer Information Bulletin CIB125, Transducer (Re)PlacementUncertainty, dated April 23, 2007 1-6 NRC Regulatory Issue Summary 2002-03, Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications, dated January 31, 2002 1-7 Instrument Society of America (ISA) S67.04 1-8 NRC Regulatory Guide 1.105, Setpoints for Safety Related Instrumentation, Revision 3, dated December 1999 1-9 ANSI/IEEE-ANS Std. 7-4.3.2. 1993, "IEEE Standard Criteria for Digital Computers in Safety Systems of Nuclear Power Generating Stations" 1-10 ASME NQA-2a-1990, "Quality Assurance Requirements for Nuclear Facility Applications" 1-11 EPRI TR-103291s, "Handbook for Verification and Validation of Digital Systems," December 1994 12 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION II. ACCIDENTS AND TRANSIENTS FOR WHICH THE EXISTING ANALYSES OF RECORD BOUND PLANT OPERATION AT THE PROPOSED INCREASED POWER LEYEL INTRODUCTION The reactor core and/or NSSS thermal power are used as inputs to most plant safety, component, and system analyses.

These analyses generally model the core and/or NSSS thermal power in one of three ways.First, some Calvert Cliffs analyses apply an explicit 2% increase to the initial condition power level to-account solely. for the power measurement uncertainty.

These analyses have not been re-performed for the requested MUR power uprate conditions because the sum of increased core power level and the decreased power measurement uncertainty falls within the previously analyzed conditions.

The power calorimetric uncertainty calculation described in Section I indicates that with the Caldon LEFM CheckPlus devices installed, the power measurement uncertainty (based on a 95% probability at a 95% confidence interval) is less than 0.6%. Therefore, these analyses only need to reflect a 0.6% power measurement uncertainty.

Currently with the RTP of 2700 MWt, the analytical power level of 2754 MWt (102% of 2700 MWt) is used in the safety analysis.

With a revised RTP of 2737 MWt and a revised uncertainty, the analytical power level is unchanged at 2754 MWt.Second, some Calvert Cliffs analyses employ a nominal initial condition power level. These analyses have been evaluated for the increased power level with the MUR power uprate. The results demonstrate that the applicable analysis acceptance criteria continue to be met at the MUR power uprate conditions.

Third, some of the Calvert Cliffs analyses are performed at zero power initial conditions or do not actually model the core power level. Consequently, these analyses have not been re-performed for the proposed MUR power uprate since they are unaffected by the core power-level.

11.1 NUCLEAR STEAM SUPPLY SYSTEM ACCIDENT EVALUATION The analyses referenced in Table 11-1 are the AOR for Calvert Cliffs Units 1 and 2. These analyses do not change, that is, they continue to remain valid for the MUR power uprate.The information in the table is organized to comply with Reference 1I-1. The first column contains the applicable Updated Final Safety Analysis Report (UFSAR) section. The second column identifies the transient, and columns three through six contain power and uncertainty information from the AOR, as well as confirmation that the AOR remains bounding with the MUR power uprate. Column seven provides the reference for the NRC's previous approval of the AOR, as well as an indication of type of approval.

Approval types are either NRC SER or performed under 10 CFR 50.59. The final column elaborates briefly on the impact of the power uprate on the AOR.The sections that follow provide details of the safety analyses.13 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION TABLE 11-1 Impact of Power Uprate on the UFSAR Chapter 14 Accident Analyses AOR ASSUMPTIONS AND REFERENCES Total ToalReference Reference UFSAR SECTION/EVENT RTP Uncert. Core Bounds Rfrn Reference (MWt) (%) Power MUR?(MWt) Approval AOR NOTES RIS 2002-03 A B,C B,C B,C B,C D D Rqmnt->14.2 Control Element 2700 +/-2 2754 Yes 11-2 11-3, 11-4 Re-analyzed for thermal margin credits seen Assembly Withdrawal with TURBO fuel. MUR has no impact.Event 14.3 Boron Dilution Event { } { _ } { -} { -} ** 11-3 Not effected by an increase RTP. Analysis based on boron concentrations and RCS volumes which are unchanged for power uprate.14.4 Excess Load Event 2700 :+/-2 2754 Yes 11-5 11-3, 11-6 Re-analyzed for thermal margin credits seen with TURBO fuel. MUR has no impact.14.5 Loss of Load Event 2700 +/-2 2754 Yes 11-2 11-7, 11-8 Evaluated for impact of MUR. Existing AOR plus uncertainty bounds the MUR total core power.14.6 Loss of Feedwater Flow 2700 +/-2 2754 Yes 11-9 11-7, 11-10 Evaluated for impact of MUR. Existing Evejnt AOR plus uncertainty bounds the MUR total core power.14.7 Excess Feedwater Heat 2700 +/-2 2754 Yes 11-5 11-3, 11-6 Re-analyzed as a sub-set of the Excess Load Removal Event event. MUR has no impact.14.8 Reactor Coolant System 2700 +/-2 2754 Yes II-5 11-7, li-11 Evaluated for impact of MtJR. Existing Depressurization AOR plus uncertainty bounds the MUR total core power.14.9 Loss-of-Coolant Flow 2700 +/-2 2755 Yes 11-2 11-3, 11-12 Re-analyzed for thermal margin credits seen Event with TURBO fuel. MUR has no impact.14.10 Loss-of-Non-Emergency 2700 +/-2 2754 Yes 11-5 11-7, 11-13 Evaluated for impact of MUR. Existing AC Power AOR plus uncertainty bounds the MUR total core power.14 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION TABLE 11-1 Tmnanct of Power linrate oin the ITFSAR Chanter 14 Accident Analyses AOR ASSUMPTIONS AND REFERENCES Total ToalReference Reference UFSAR SECTION/EVENT RTP Uncert. Core Bounds Rfrn Reference (MWt) (%) Power MUR? atNC 5.9 (MWt) Power Approval AOR NOTES RIS 2002-03 A B,C B,C B,C B,C D D Rqmnt--->14.11 Control Element 2700 +/-2 2754 Yes 11-5 11-3, 11-14 Re-analyzed for thermal margin credits seen Assembly Drop Event with TURBO fuel. MUR has no impact.14.12 Asymmetric Steam 2700 +/-2 2754 Yes 11-5 11-3, 11-7, Re-analyzed for thermal margin credits seen Generator Event 11-15 with TURBO fuel. MUR has no impact.14.13 Control Element 2700 +/-2 2754 Yes II-5 11-3, 11-16 Evaluated for impact of MUR. Existing Assembly Ejection AOR plus uncertainty bounds the MUR total core power.14.14 Steam Line Break Event 2700 +/-2 2754 Yes 11-2 11-3, 1-15, Pre-trip portion re-analyzed for thermal 11-17, 11-18, margin credits seen with TURBO fuel. Post-11-19, 11-20, trip re-analyzed for cycle specific credits.11-21 MUR has no impact on either portion of the event.14.15 Steam Generator Tube 2700 +/-2 2754 Yes 11-2 11-15, 11-22 Evaluated for impact of MUR. Existing Rupture Event AOR plus uncertainty bounds the MUR total core power.14.16 Seized Rotor Event 2700 +/-2 2754 Yes II-5 11-3, 11-23 Re-analyzed for thermal margin credits seen with TURBO fuel. MUR has no impact.14.17 Loss-of-Coolant Accident 2700 +2 2754 Yes 11-24 11-3, 11-7, Evaluated for impact of MUR. Existing 11-25, 11-26 AOR plus uncertainty bounds the MUR total core power.14.17.2 Large Break LOCA 2700 +/-2 2754 Yes 11-24 11-3, 11-7, 11-26 14.17.3 Small Break LOCA 2700 +2 2754 Yes 11-24 11-3, 11-7, 11-25 15 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION TABLE 11-1 Impact of Power Uprate on the UFSAR Chapter 14 Accident Analyses AOR ASSUMPTIONS AND REFERENCES Total ToalReference Reference UFSAR SECTION/EVENT RTP Uncert. Core Bounds R c Reernc (MWt) (%) Power MUR? L as N0. NOTES (MWt) Approval AOR RIS 2002-03 A B,C B,C B,C B,C D D Rqmnt--)14.18 Fuel Handling Incident 2700 +2 2754 Yes ** 11-27, 11-28 Evaluated for impact of MUR. Radionuclide inventories based upon 2754 MWt. Existing AOR plus uncertainty bounds the MUR total core power.5.3.1.2. Turbine-Generator

{ _ } { -} { _} { } ** 11-29, 11-30 Not effected by power increase.

Analysis Overspeed Incident based on pitching turbine blades.14.20 Containment Response 2700 +2 2754 Yes ** 11-31 Evaluated for impact of MUR. Existing AOR plus uncertainty bounds the MUR total core power.14.21 Hydrogen Accumulation

{ _ } { _ } { } { -} ** No Longer A change to the Calvert Cliffs Technical in Containment Analyzed Specifications removed this incident.for Chapter 14 14.22 Waste Gas Incident 2700 +2 2754 Yes ** 11-32 Evaluated for impact of MUR. Radionuclide inventories based upon 2754 MWt. Existing AOR plus uncertainty bounds the MUR total core power.14.23 Waste Processing System 2700 +/-2 2754 Yes ** 11-32 Evaluated for impact of MUR. Radionuclide Incident inventories based upon 2754 MWt. Existing AOR plus uncertainty bounds the MUR total core power.14.24 Maximum Hypothetical 2700 +2 2754 Yes ** 11-33 Evaluated for impact of MUR. Radionuclide Accident inventories based upon 2754 MWt. Existing AOR plus uncertainty bounds the MUR total core power.16 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION TABLE II-1 Impact of Power Uprate on the UFSAR Chapter 14 Accident Analyses AOR ASSUMPTIONS AND REFERENCES Total UFSAR SECTION/EVENT RTP Uncert. Core Bounds Reference Reference (MWt) (%) Power MUR? Last NRC 50.59/(MWt) Approval AOR NOTES RIlS 2002-03 A B,C B,C B,C B,C D D Rqmnt-)14.25 Excessive Charging { -} { _ } { _} { } ** 1-15, 11-34 Not affected by power increase.

Evaluated to Event assure that the operator has at least 15 minutes from initiation of high pressure level alarm to take corrective action and terminate the event prior to filling the pressurizer solid.14.26 Feed line Break Event 2700 +/-2 2754 Yes 11-2 11-3, 11-35 Evaluated for impact of MUR. Existing AOR plus uncertainty bounds the MUR total core power.** -Not applicable for reference to previous NRC review 17 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION 11.2 NON-LOSS-OF-COOLANT ACCIDENT/TRANSIENT ANALYSES All of the UFSAR Chapter 14 non-LOCA transient analyses were evaluated for increase in RTP due to the MUR power uprate. The analyses include the NSSS response with replacement SGs (References 11-7 and II-11). Replacement SGs decreased the number of plugged SG tubes, which in turn increased RCS flow. For some events, the original SG results are reported in the UFSAR because they are representative of the replacement SG results.Many of the events were reanalyzed for thermal margin credit associated with TURBO fuel and these events included a target power uprate of 2746 MWt (1.7%), which bounds the proposed change in RTP to 2737 MWt. The events that use the target of 2746 MWt include a 0.3% uncertainty.

The uprated RTP with uncertainty is equivalent to the pre-uprate total core power, which is 2754 MWt.In the evaluation of the remaining events (those not reanalyzed for TURBO fuel), the existing assumption on core power plus uncertainty bounds the MUR power uprate. For all events, no changes to the Reactor Protective System (RPS) or Engineering Safety Features were assumed or were necessary.

The evaluation of the UFSAR Chapter 14 non-LOCA transient analyses concludes that the current analyses are applicable for Calvert Cliffs with the MUR power uprate.11.2.1 Control Element Assembly Withdrawal Event (UFSAR 14.2)A failure in either the Control Element Assembly (CEA) Drive Mechanism Control System or the Reactor Regulating System may initiate a sequential bank withdrawal, inserting positive reactivity.

and causing increases in reactor power, RCS temperature, and RCS pressure.

The eyent is terminated by either the Variable High Power Trip, the High Pressurizer Pressure Trip, the Thermal Margin/Low Pressure (TM/LP) Trip, or the insertion of negative reactivity due to Doppler and negative Moderator Temperature Coefficient (MTC) feedbacks.

The current AOR for the CEA Withdrawal Event is analyzed and documented in Reference 11-4. In support of the MUR power uprate, this referenced analysis is performed with the assumption of a rated power of 2746 MWt plus uncertainties, which bounds the MUR power uprate power level~of 2737 MWt plus uncertainties.

This re-analysis also implements the Asea Brown Boveri, Inc.-Turbo Vane (ABB-TV)correlation for critical heat flux (approved in Reference 11-36), and makes all appropriate input and assumption adjustments associated with both ABB-TV and the MUR power uprate. Approved methodologies and codes (References 11-2, 11-37, 11-38, and 11-39) were used, along with approved associated limits/constraints and acceptance criteria.

As with all applicable UFSAR Chapter 14 analyses, associated with implementation of the ABB-TV critical heat flux correlation was a change in the departure from nucleate boiling (DNB) specified acceptable fuel design limits (SAFDL) to a value of 1.24, determined by application of extended statistical combination of uncertainties (Reference 11-40).This value is acceptable in relation to the NRC-approved minimum departure from nucleate boiling ratio (DNBR) value of 1.13 associated with the approved methodologies of this analysis.

Assuming a rated full power level of 2746 MWt plus uncertainties and implementing ABB-TV, all acceptance criteria are met with respect to DNBR, peak linear heat generation rate (PLHGR), maximum primary and secondary pressure and radiological consequence.

The analysis for CEA withdrawal is acceptable relative to applicable SERs and bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.2 Boron Dilution Event (UFSAR 14.3)A Boron Dilution Event is defined as any event caused by a malfunction or an inadvertent operation of the Chemical and Volume Control System (CVCS) that results in a dilution of the active portion of the 18 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION RCS. The analysis of this event covers all six modes of operation, each mode being associated with a required minimum time to lose required shutdown margin. This analysis, most recently documented in Reference 11-3 for the current operating conditions, is unaffected by the proposed MlUR power uprate.The analysis is based on RCS and CVCS volumes, along with the boron concentration, to show that operator action within the required minimum time period will terminate the dilution prior to violating the assumed parameters for shutdown margin. The boron dilution event assumes boron concentration levels associated with operating modes, which continues to bound the MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.3 Excess Load Event (UFSAR 14.4)An Excess Load Event, as documented in the Calvert Cliffs UFSAR, is a rapid uncontrolled increase in SG steam flow not caused by a Steam Line Break (SLB). In the assumed presence of a negative MTC and Fuel Temperature Coefficient, positive reactivity addition leads to an increase in core power level, decreasing DNBR and linear heat rate (LHR) margin. The transient continues until the Variable High Power Trip is reached on neutron flux or core temperature differential (AT), terminating the event. The limiting scenario is most likely to be caused by a full opening of the turbine control valves, atmospheric dump valves (ADVs), or turbine bypass valves during steady-state operation.

Limiting cases are determined at both Hot Full Power (HFP) and Hot Zero Power (HZP).The current AOR, as documented in Reference 11-6, bounds operation at the MUR power uprate power level of 2737 MWt plus uncertainties.

That AOR also has been verified to use approved methodologies and codes, along with all associated limits and conditions as prescribed by associated SERs (References 11-5, 11-36, and 11-39). The current AOR at HFP assumes an initial reactor thermal power of 2754.2 MWt, including uncertainties.

This thermal power level bounds the proposed MUR power uprate power level of 2737 MWt, plus uncertainties.

All criteria for acceptance are met with respect to DNBR, PLHGR, pressure limits, and radiological consequence.

The Excess Load Event AOR bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.4 Loss of Load Event (UFSAR 14.5)As defined in UFSAR Section 14.5, a Loss of Load Event is defined as any event that results in a reduction in the SGs' heat removal capacity through a loss of secondary steam flow. Such an event could be caused by a closure of all main steam isolation valves (MSIVs), turbine stop valves, or turbine control valves along the steam flow path between the SGs and the high pressure turbine. The most limiting Loss of Load Event is a turbine trip without concurrent reactor trip, or an inadvertent closure of the turbine stop valves at HFP.Reference 11-8 provides a bounding AOR for both Calvert Cliffs Units 1 and 2. The assumed power level for transient initiation at HFP is 2771 MWt, which includes a 2.0% instrument uncertainty and a conservative assumption of an additional 17 MWt for RCP heat. This assumed power level in the analysis of 2771 MWt bounds the proposed operation at an MUR power uprate power level of 2737 MWt and the power measurement uncertainty.

All assumptions and methodologies associated with and documented in the AOR are consistent with previously approved analyses and associated SERs and limitations/conditions for application (Reference 11-39 for CESEC-III).

All acceptance criteria were found to be met for the bounding analysis with respect to DNBR, fuel performance, peak pressures (RCS and secondary), and radiological consequence.

This analysis, having been performed at HFP with a thermal power of 2771 MWt (including uncertainty and RCP heat), bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION 11.2.5 Loss of Feedwater Flow Event (UFSAR 14.6)A Loss of Feedwater Flow Event is defined as a reduction or loss of feedwater to the SGs without a corresponding reduction in steam flow from the SGs. The most limiting Loss of Feedwater initiating event is determined to be an inadvertent instantaneous closure of the feedwater regulating valves, which results in the largest steam and feedwater flow mismatch and the most rapid reduction in SQ inventory.

The transient causes an increase in primary and secondary pressures and is ultimately terminated by the High Pressurizer Pressure Trip or the Low SG Level Trip to ensure that all acceptance criteria are met.The current AOR for this event was documented in References 11-7 and 11-10 and was reviewed and accepted by the NRC as documented in Reference 11-9. The assumed initial power for the transient is 2771 MWt (including uncertainties and RCP energy), which bounds operation at the MUR power uprate power level of 2737 MWt plus uncertainties.

All acceptance criteria were met with respect to DNBR, PLHGR, peak pressures, SG inventory, and radiological consequences.

The current AOR for Loss of Feedwater Flow bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.6 Excess Feedwater Heat Removal Event (UFSAR 14.7)The Excess Feedwater Heat Removal Event results from an extraction of excessive heat from the RCS through the SGs caused by a reduction in S6 feedwater temperature without a corresponding reduction in steam flow from the SGs. The limiting circumstance of a loss of both high pressure feedwater heaters, coupled with the presence of a conservatively negative MTC and Fuel Temperature Coefficient, results in a core power increase due to the corresponding decrease in RCS temperature.

This reactor power increase causes the system to approach the SAFDLs, and is ultimately mitigated by the Variable High Power Trip.This analysis is documented as an Appendix to the Excess Load analysis of Reference 11-6 and discussed in Reference 11-3. This analysis is bounded by the inputs and results of the AOR for the Excess Load Event. As the Excess Load Event has already been determined to bound operation at the proposed MUR power uprate power level of 2737 MWt, the Excess Feedwater Heat Removal Event is also bounded by the current AOR. Bounding inputs with respect to initial reactor core power level (2754.2 MWt, including uncertainties), and associated methodologies, are identical to those discussed for the Excess Load Event. As such, the current AOR for the Excess Feedwater Heat Removal Event bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.7 Reactor Coolant System Depressurization Event (UFSAR 14.8)The RCS Depressurization Event is considered an Anticipated Operational Occurrence (AOO) for which action of the RPS is required to prevent SAFDL violation.

The event is initiated by assuming the inadvertent opening of both power-operated relief valves, resulting in a rapid depressurization of the RCS.The analysis shows that action of the RPS by way of the TM/LP trip prevents exceeding the associated SAFDLs, particularly DNBR.As stated in the AOR, the assumed initial core power does not affect the results of the event. However, the documented AOR (Reference I-11, justifying results of Reference 11-41 with replacement SGs) is performed with an assumed initial reactor core power level of 2771 MWt (including uncertainties and RCP energy), which bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

20 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION 11.2.8 Loss of Coolant Flow Event (UFSAR 14.9)The Loss of Coolant Flow Event is classified as an AOO for which RPS trips and/or sufficient initial steady-state thermal margin, maintained by the applicable Technical Specifications, are necessary to prevent exceeding acceptable limits. This transient event is initiated from a HFP condition and modeled to envelope the occurrence of two separate postulated scenarios for losing power to the RCPs: a complete loss of alternating current (AC) to the plant, and a failure of the fast transfer breakers to close following an assumed trip of the main generator.

The intermediate system response to the RCP coast down is a rapid decrease in coolant mass flow rate through the reactor core, causing a rise in enthalpy across the core in the direction of coolant flow. A relatively slight power increase results, due to the assumed presence of a positive MTC. The main concern with respect to SAFDLs for this event is DNBR, which is met in the analyses (Reference 11-12) by ensuring that initial steady-state margin is built into the DNB design operating limit such that, in conjunction with crediting of the low flow trip function, the DNBR SAFDL is not exceeded.The Loss of Coolant Flow Event AOR (Reference 11-12) credits the thermal margin gains associated with implementation of TURBO fuel. For this analysis, the maximum core power with uncertainties applied was 2755 MWt, (2746 MWt plus uncertainties).

The analyzed maximum power level of 2755 MWt bounds the proposed MUR poweruprate power level of 2737 MWt, plus uncertainties and rounded up.Methodologies associated with this analysis were verified to be consistent and within the limitations and conditions of associated SERs (References 11-5, 11-36, 11-42, 11-43, and 11-44) and previously NRC-approved analyses (Reference 11-2).All acceptance criteria were met with respect to DNBR, PLHGR, peak pressures, and radiological consequence.

The maximum analyzed power level of 2755 MWt (including uncertainties) and assumed RTP of 2746 MWt bound operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.9 Loss of Non-Emergency AC Power Event (UFSAR 14.10)The Loss of Non-Emergency AC Power Event involves a loss of electrical power to RCPs, resulting in an RCS flow coast down that challenges SAFDLs and yields an increased steam release to the atmosphere via the main steam safety valves (MSSVs) and ADVs. With respect to DNBR and PLHGR, this event is bounded by the Loss of Coolant Flow Event described above and documented in References 11-7 and 11-12. Loss of Coolant Flow has been verified to bound operation at the MUR power uprate power level, with use of applicable approved codes, methodologies, and limitations/constraints.

The Loss of Non-Emergency AC Power was evaluated and documented as an AOR in Reference 11-13.An explicit analytical calculation was not performed for the reanalysis, but the documented AOR justifies the results of the previous AOR for operation with the replacement SGs. The analysis was performed at an initial power level of 2754 MWt, including uncertainties, which bounds the proposed MUR power uprate power level of 2737 MWt, plus uncertainties.

As previously stated, all SAFDL limits, including DNBR, are bounded by the Loss of Coolant Flow Event. Additionally, the peak pressures associated with Loss of Non-Emergency AC are bounded by the results of the Loss of Load Event (also discussed above).Results of the AOR (Reference 11-13) meet all applicable criteria and are verified to be produced by NRC-approved methodologies in accordance with applicable SERs. Current analysis for the Loss of Non-Emergency AC Power bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION 11.2.10 Control Element Assembly Drop Event (UFSAR 14.11)The CEA Drop Event entails the drop of a single full length CEA into the core, reducing fission power in the vicinity of the dropped CEA and adding negative reactivity core-wide.

A prompt drop in core power and heat flux results from the negative reactivity insertion, the magnitude of which depends on the reactivity worth of the dropped CEA. Assuming an inoperable turbine runback circuit, the resulting power mismatch between the primary and secondary systems leads to a cooldown of the RCS and a subsequent positive reactivity addition due to the effects of a negative MTC. Doppler reactivity and moderator feedbacks ultimately terminate the reactivity excursion, producing a re-stabilized core condition with an asymmetric power distribution and correspondingly higher peaking factors. Criteria with respect to DNB, PLHGR and radiological consequence must be shown analytically to be met.A new AOR was established for both Units 1 and 2 with References 11-3 and 11-14. The AOR implements the methodologies associated with TURBO fuel and ensures bounding inputs and results for the anticipated MUR power uprate. Rated power for this event is assumed to be 2746 MWt, and the maximum initial power including uncertainties is 2754.2 MWt, which bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

The performance of this analysis has been verified by the vendor and Calvert Cliffs to have been done in accordance with all applicable SERs (References 11-5, 11-36, and 11-39) and limitations/conditions.

All results are shown to be acceptable with respect to the acceptance criteria for DNBR, PLHGR, peak pressures, and radiological consequence.

The current AOR for the CEA Drop Event bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.11 Asymmetric Steam Generator.

Event (UFSAR 14.12)The Asymmetric SG Event is classified as an AOO, described as a rapid imbalance in heat transfer between the two SGs, initiated by one of the following:

a loss of load to one SG, excessive increase in load to one SG, loss of feedwater to one SG, or excessive feedwater flow increase to one SG. The limiting cause evaluated for the current AOR at Calvert Cliffs is a loss of load to one SG, caused by instantaneous closure of one of two MSIVs. This circumstance produces the most rapid temperature tilt across the core, resulting in a limiting approach to the DNBR SAFDL for this analysis.The current bounding AOR for Units 1 and 2 is documented in Reference 11-45. This revision to the AOR explicitly addresses the implementation of TURBO fuel, ABBnTV critical heat flux correlation, and the MUR power uprate. Rated power for this analysis is assumed to be 2746 MWt, and the maximum initial power, including uncertainties, for the analysis is assumed to be 2754.2 MWt, Methodologies and codes associated with this analysis are verified to be consistent and within the limitations and conditions of associated SERs (References 11-5, 11-36, and 11-39) and previously NRC-approved analyses (Reference 11-5). All acceptance criteria were met with respect to DNBR, PLHGR, peak pressures, and radiological consequence.

As such, the current AOR for the Asymmetric SG Event bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.12 Control Element Assembly Ejection Event (UFSAR 14.13)The CEA Ejection Event results from a postulated complete circumferential break of the control element drive mechanism (CEDM) housing or of the CEDM nozzle on the RV head. The analysis is performed from postulated HFP and HZP initial conditions, each resulting in a rapid core power increase for a brief period of time. Doppler reactivity feedback inhibits the core reactivity and power rise, and the reactor is ultimately shutdown by a high power level trip, thereby terminating the transient.

The core is protected from fuel damage by CEA insertion limits associated with various power levels (Power-Dependent Insertion Limit of the Technical Specifications) and the high power trip. Being a postulated event, a 22 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION small fraction of fuel failure is permitted in the analysis within the restrictions of criteria for acceptance placed on deposited energy limits and offsite radiological consequence.

The current bounding AOR for Units I and 2 at Calvert Cliffs is documented in Reference 11-16. This revision to the AOR explicitly addresses the implementation of Zirconium Diboride (ZrB 2) fuel with axial blankets, as well as ZIRLOTM cladding and encompasses the MUR power uprate. Rated power for this analysis is assumed to be 2746 MWt, and the maximum initial power, including uncertainties, for the analysis is assumed to be 2754 MWt. Methodologies and codes associated with this analysis are verified to be consistent and within the limitations and conditions of associated SERs (References 11-42, 11-46, 11-47, and 11-48) and the NRC-approved analysis (Reference 11-5). All acceptance criteria were met with respect to fuel clad failure and radiological consequence.

As such, the current bounding AOR for the CEA Ejection Event bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.13 Steam Line Break Event (UFSAR 14.14) I A SLB Event is defined as a breach in the Main Steam piping that carries steam from the SGs to the turbine-generator and other equipment.

That breach in the main steam piping produces an increase in heat extraction by the SGs, causing a cooldown of the RCS. In the presence of an assumed negative MTC, that RCS cooldown leads to an addition of positive reactivity to the RCS. The transient is terminated by a reactor trip associated with the severe decrease in SG pressure, and the MSTVs in the main steam line close to isolate steam flow from the affected SG. The SLB Event is divided analytically into two separate phases, pre-trip and post-trip for separate safety concerns and associated evaluation against respective acceptance criteria.

The primary concern in the pre-trip SLB analysis is the power excursion related to the RCS cooldown and the assumed negative MTC. A loss of power coincident with reactor trip is also assumed. A limiting combination of break size and MTC is determined parametrically for SLBs both inside and outside Containment during the pre-trip SLB analysis.

The primary concern associated with the post-trip analysis is a return-to-power in the vicinity of an assumed stuck control rod.Limiting scenarios with respect to DNBR are determined parametrically for HFP and HZP initial conditions, both with and without loss of power.The critical heat flux correlation utilized in the SLB analysis is the MACBETH correlation, NRC-approved in Reference 11-49. Associated with that documented SER is a minimum DNBR limit of 1.30 for the MACBETH critical heat flux correlation.

Additional applicable SERs for the SLB and this discussion are References 11-39 and 11-50.The current AOR for the pre-trip SLB was established in References I1-15 and 11-19. The maximum initial power level at event initiation assumed in that analysis is 2754.2 MWt including uncertainties, which bounds the proposed MUR power uprate power level of 2737 MWt (plus uncertainties) at Calvert Cliffs. This AOR for the pre-trip SLB credits the thermal margin gains associated with TURBO fuel and the ABB-TV critical heat flux correlation, and bounds the proposed MUR power uprate operation.

Acceptance criteria with respect to DNBR, PLHGR, peak pressures and radiological consequence are all met. The current AOR for the pre-trip SLB bounds operation at the proposed MUR power.uprate power level of 2737 MWt plus uncertainties.

The post-trip SLB Event is currently analyzed separately for each operating cycle to credit the cycle-specific physics input to the analysis.

The current AORs for Units I and 2 are documented in References 11-20 and 11-21, respectively.

The AORs for the two operating units employ the MACBETH critical heat flux correlation (design DNBR _ 1.30, SER Reference 11-49). Each current AOR for the post-trip SLB is also performed with an assumed rated power level of 2746 MWt and a maximum initial total power of 2754 MWt including uncertainties.

This power level input assumption bounds the 23 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION proposed operation at the MUR power uprate power level of 2737 MWt. The AOR for Units 1 and 2 (References 11-20 and 11-21) also bound operation with ZrB 2 Integral Fuel Burnable Absorber (IFBA) in conjunction with axial blankets.

All applicable restrictions, limits and conditions associated with the respective methodologies, codes, and correlations are met within the bounds of respective appropriate SER references.

The current AOR for the post-trip SLB bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.14 Steam Generator Tube Rupture Event (UFSAR 14.15)The SG Tube Rupture Event is a breach of the barrier between the RCS and the main steam system (MSS), resulting in mass transfer between the primary and secondary systems and, more consequentially, a radiological release to the environment through the MSSVs and the ADVs.The current bounding AOR for the SG Tube Rupture Event is documented in Reference 11-22. As the primary concern associated with this analysis is radiological consequence, a reanalysis was not explicitly performed for TURBO fuel implementation.

The AOR is documented as bounding in terms of affected neutronic parameters (e.g., Scram curves) for implementation of ZrB 2 IFBAs in conjunction with axial blankets (Reference 11-42). The AOR is also supported by Reference 11-51 with regard to justifying parameter assumptions related to proportional and backup heater nominal heat rates, MSSV setpoints, and charging pump flow. The assumed maximum power level at initiation of the transient from HFP conditions in the AOR is 2754 MWt, which bounds operation at the MUR power uprate power level of 2737 MWt, plus uncertainties.

The SG Tube Rupture event as documented in the current AOR bounds operation at the proposed MUR power uprate power level including uncertainties, and meets the requirements, limitations and conditions associated with all applicable SERs.11.2.15 Seized Rotor Event (UFSAR 14.16)The Seized Rotor Event is classified as a postulated event, for which a limited amount of fuel failure is permitted within the bounds of associated acceptance criteria.

The transient event is caused by an instantaneous seizure of a RCP shaft, postulated to occur as a result of mechanical failure or a loss of component cooling water to the RCP shaft seals. The flow rate rapidly reduces to a value corresponding to three RCPs, as opposed to four. The corresponding reduction in RCS flow rate causes a reactor trip on low RCS flow. The reduction of RCS flow rate results in a degradation of DNBR with respect to the SAFDL.Reference 11-23 documents the current AOR for the Seized Rotor Event. The AOR credits the thermal margin benefits of TURBO fuel, as realized by application of the ABB-TV critical heat flux correlation in conjunction with CETOP-D (References 11-36 and 11-38). The effects of implementation of ZrB 2 fuel in conjunction with, axial blankets are evaluated in Reference 11-52. The assumed maximum power level for the currently bounding AOR is 2754 MWt, which bounds the proposed operation following MUR power uprate of 2737 MWt, plus uncertainties.

As all acceptance criteria with respect to DNBR, PLHGR, peak pressures, and radiological consequence are met within the restrictions, limitations, and constraints of NRC-approved methodologies and codes, the Seized Rotor Event as currently analyzed bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.16 Fuel Handling Incident (UFSAR 14.18)The Fuel Handling Incident analysis assumes that a fuel assembly is dropped during fuel handling, either in the Containment or in the Spent Fuel Pool. The results of this analysis are dependent upon the radionuclide inventory assumed for the dropped fuel assembly.

The inventories associated with this analysis have been generated based on an assumption of core operating power of 2754 MWt, and source term values are based on the TID-14844 methodology in accordance with Regulatory Guide 1.25.24 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION Therefore, the current analysis relating to the Fuel Handling Incident bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

Reference 11-66 approves Technical Specification changes requested in Reference 11-67 associated with the implementation of the alternative radiological source term (AST). The AST methodology replaces the existing accident radiological source term that is described in TID-14844.

The Fuel Handling Incident was reanalyzed using AST and the AOR, documented in Reference 11-68, assumed the core isotopic inventory is based upon a maximum full power operation of 254 MWt. Calvert Cliffs expects to switch to the AST methodology for the Fuel Handling Incident in the year 2010, and since the reanalysis of the Fuel Handling Incident was performed assuming operation at 2754 MWt, the reanalysis bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.17 Turbine-Generator Overspeed Incident (UFSAR 5.3.1.2)The Turbine-Generator Overspeed Incident is an analyzed event based on the failure of rotating elements of the steam-turbines and generators.

This analysis is not a Design Basis Event (DBE) or AOO and is documented in detail in UFSAR Section 5.3.1.2. The thermal power increase related to the MUR power uprate does not impact the results of this analysis.

As such, the Turbine-Generator Overspeed Incident bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.18 Hydrogen Accumulation in Containment This analysis has been deleted from the UFSAR per License Amendment Nos. 262/239. Reanalysis is not required to verify that the analysis bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.19 Waste Gas Incident (UFSAR 14.22)The limiting Waste Gas Incident analyzed for UFSAR Chapter 14 is an uncontrolled and unexpected release to the atmosphere of radioactive xenon and krypton fission gases stored in one waste decay tank.The assumed maximum activity, in accordance with Reference 11-32, is determined based on conditions in the waste gas decay tank shortly after plant heatup and startup after cold shutdown conditions near the end of a 24-month operating cycle. Associated limiting activity levels are calculated in Reference 11-32 with the assumption of constant full-power operation at 2754 MWt. Radiological consequence limits are met. The Waste Gas Incident bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.20 Waste Processing System Incident (UFSAR 14.23)The Waste Processing System Incident assumes a seismically-induced failure of the reactor coolant Waste Processing System whereby the contents of the system are released.

Reference 11-32, as discussed in Section 11.2.19, contains the analysis for this event. As previously mentioned, the depletion calculations for generating radio-isotopic inventories for these analyses is performed at a core thermal power level of 2754 MWt. Therefore, the Waste Processing System Incident analysis bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.2.21 Maximum Hypothetical Accident (UFSAR 14.24)The results of this analysis demonstrate bounding compliance with the guidelines of 10 CFR Part 100. As stated in UFSAR Section 14.24, the pre-accident thermal power for the Maximum Hypothetical Accident is 2754 MWt. All methodologies and results are consistent with approved methodologies and previously submitted analyses.

The documented Maximum Hypothetical Accident bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

25 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION 11.2.22 Excessive Charging Event (UFSAR 14.25)The Excessive Charging Event is analyzed to verify compliance with the limits of Technical Specification 3.4.4, and to provide the basis for associated alarm setpoints.

Specifically, the AOR, Reference 11-34, verifies that operator action no sooner than 15 minutes following receipt of pressurizer high level alarm suffices to terminate the event without violating limits on pressurizer level. The associated analysis is based on RCS volumes and CVCS flow rates (letdown and charging).

Reactor power level does not affect the results. The current AOR is bounding and acceptable with respect to the plant configuration (charging pump flows, installed pressurizer level setpoints, etc.), and remains valid for the power level associated with the proposed MUR power uprate, including uncertainties.

11.2.23 Feedline Break Event (UFSAR 14.26)The Feedline Break Event is a postulated accident whereby a piping failure occurs downstream of the check valves between the SG and Containment.

The affected SG empties, causing elevated temperatures in that SG and the RCS. A reactor trip occurs on either loss of SG Level or High Pressurizer Pressure, terminating the pressure transient in combination with the opening action of the pressurizer safety valves and MSSVs.The AOR for the Feedline Break Event is contained in Reference 1I-35,-and described in Reference 11-52, and bounds operation under current and proposed MUR power uprate power levels. The maximum core power level assumed in the analysis as an input condition is .2771 MWt, including rated power plus uncertainties and RCP energy. All acceptance criteria for the event with regard to DNBR, peak RCS and secondary pressure limits, radiological consequence, and long-term cooling capability are verified to have been met. Additionally, all methodologiesand code implementation are consistent with the most recent NRC-reviewed analysis documented in Reference 11-2. Compliance with applicable SERs is verified for use of CESEC-II1 (Reference 11-39). As all results and methodologies are acceptable, all results meet associated acceptance criteria, and the maximum initial power level exceeds the proposed MUR power uprate plus uncertainties, the results of the current Feedline Break Event AOR bound the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.3 EMERGENCY CORE COOLING SYSTEM PERFORMANCE The Calvert Cliffs Units 1 and 2 ECCS performance analysis consists of a large break loss-of-coolant accident (LBLOCA) and a small break loss-of-coolant accident (SBLOCA) analysis.

Both analyses were performed at a core power level of 2754 MWt. Consistent with the original requirement of Paragraph I.A of Appendix K to 10 CFR Part 50, 2754 MWt is equal to 102% of the current licensed core power level, i.e., RTP of 2700 MWt.The Calvert Cliffs Units 1 and 2 LBLOCA and SBLOCA analyses were performed with the 1999 Evaluation Model (EM) (Reference 11-53) and Supplement 2 to CENPD-137 Evaluation Model (S2M)(Reference 11-54) versions of the Westinghouse ECCS EMs for CE pressurized water reactors (PWRs).The SERs for the 1999 EM (Reference 11-55) and the S2M (Reference 11-56) generically approved the EMs for referencing in licensing applications for CE designed PWRs. The two EMs were specifically accepted for Calvert Cliffs Units 1 and 2 as allowed analytical methods for use in determining core operating limits in Reference 11-57. A summary of the Calvert Cliffs LBLOCA and SBLOCA analyses using the 1999 EM and the S2M was provided to the NRC in Reference 11-58. Detailed descriptions of the analyses are contained in Calvert Cliffs UFSAR Section 14.17.As allowed by Paragraph I.A of Appendix K, Calvert Cliffs Nuclear Power Plant proposes to increase the licensed core power level and decrease the power measurement uncertainty such that the analytical core 26 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION power level, after accounting for the new power measurement uncertainty, remains equal to 2754 MWt.Since the Calvert Cliffs Units I and 2 ECCS performance analysis was performed at an analytical core power level of 2754 MWt, it complies with Paragraph I.A of Appendix K for the proposed values for the licensed corepower level and power measurement uncertainty.

A review of the impact that the proposed increase in licensed core power level (2737 MWt) has on the Calvert Cliffs Unit I values for plant data used in the Calvert Cliffs Units 1 and 2 ECCS performance analyses concluded that the increase in power does not affect the applicability of the analysis to Calvert Cliffs Unit 1 under the MUR power uprate conditions.

The analyses and evaluations were performed for Calvert Cliffs Units I and 2. In some cases where cycle specific data is needed the analyses/evaluations targeted Unit 1 as the lead unit for the MUR power uprate. Consequently, for Calvert Cliffs Unit 1, there are no changes to the peak cladding temperature or any other result of the Calvert Cliffs Units 1 and 2 ECCS performance analyses as a consequence of the proposed changes to the licensed core power level and power measurement uncertainty.

Confirmation of the applicability of the analyses and evaluations on future cycles of Unit 2, and subsequent cycles of Unit 1, will be performed as part of the normal reload design process.The 1999 EM and the S2M EMs consist, in part, of topical reports that were written prior to the revision to Paragraph I.A of Appendix K. Some of those earlier topical reports contain statements that the analyses will use 102% of the licensed core power level. For example,Section III.A of CENPD-132P (Reference 11-59) states that "The reactor will be assumed to be operating at a power level of 102% of the maximum licensed power." Subsequent to the revision to Paragraph L.A of Appendix K, the topical reports that comprise the LBLOCA and SBLOCA EMs were not amended to reflect the revision to Appendix K; i.e., sentences like the above were not revised. As identified in the Introduction Section, Calvert Cliffs requests that approval of this license amendment request constitutes approval to apply the EMs at the proposed core power level and power measurement uncertainty.

11.3.1 Loss-of-Coolant Accident (UFSAR 14.17)The LOCA Analyses are performed in order to provide confirmation of the ECCS performance within the criteria listed in 10 CFR 50.46. The following two subsections address the AOR for both large break and small break LOCA with respect to the projected MUR power uprate power level of 2737 MWt.11.3.1,1 LBLOCA The current AOR bounding operation for Units I and 2, are found in Reference 11-26. The methodology was generically approved by the NRC and documented in Reference 11-24. The results of that analysis are applicable to the following plant configuration conditions:

• RTP (including measurement uncertainty)

< 2754 MWt* Maximum integrated radial peaking factor, Fr, max, Core Operating Limits Report (COLR) limit of 1.65 (full power, all rods out operation)" Full core representation of the TURBO fuel assembly design* Value added fuel (VAP), ZIRLOTM clad, ZrB 2 IFBA, and U0 2 fuel rod designs operating at a PLHGR of 14.5 kw/ft with 2x6-inch low-enriched axial blankets with annular pellets* Once-burned VAP ZIRLOTM clad Erbia fuel rod designs operating at 14.0 kW/ft PLHGR* SGs with < 10% tube plugging This bounding analysis employs the "1999 EM" version of Westinghouse's LBLOCA ECCS Performance Evaluation Model for Combustion Engineering designed Pressurized Water Reactors 27 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION (PWRs), as documented in Reference 11-53 and NRC-approved in Reference 11-55, conforming to the requirements associated with ZIRLOTM (SER Reference 11-46) and ZrB 2 (SER Reference 11-42).The ECCS acceptance criteria of 10 CFR 50.46 are compared to the calculated results for the bounding LBLOCA *analysis for any Calvert Cliffs operating cycle that meets the aforementioned applicability criteria.Parameter Criterion Result Peak Cladding Temperature

< 2200OF 2057°F Maximum Cladding Oxidation

< 17% 9.95%Maximum Core-Wide Oxidation 1% < 0.99%Coolable Geometry Yes Yes All results for the bounding LBLOCA analysis are acceptable with respect to acceptance criteria applied by 10 CFR 50.46. The LBLOCA, as evinced by the foregoing discussion, is performed according to all applicable SERs and bounds operation at the proposed MUR power uprate power level of 2737 MWt plus uncertainties.

11.3.1.2 SBLOCA The current AOR for SBLOCA applicable to Units I and 2 and future applicable Calvert Cliffs operating cycles .is documented in Reference 11-25 and discussed in Reference 11-3. The results of Reference 11-25 are applicable to the following plant configuration conditions:

• RTP (including measurement uncertainty)

< 2754 MWt* TURBO fuel assembly design* VAP, ZIRLO T M and Zircaloy-4 clad U0 2 fuel, with and without Erbia IFBA* VAP, ZIRLOTM clad, ZrB 2 IFBA, and U0 2 fuel rod designs with 2x6-inch low-enriched axial-blankets with annular pellets" SGs with < 10% tube plugging" PLHGR of 15.0 kW/ft This SBLOCA ECCS performance analysis is performed with the NRC-accepted S2M version of the Westinghouse CE SBLOCA EM (Reference 11-56). As documented above for the LBLOCA for both units, the bounding AOR for SBLOCA complies with all limitations and conditions of applicable SERs, such as those associated with ZIRLOTM and ZrB 2.The results demonstrate conformance for a bounding SBLOCA analysis (within the conditions of applicability) with respect to acceptance criteria of 10 CFR 50.46 as follows.Parameter Criterion Result Peak Cladding Temperature

< 2200°F 1855 0 F Maximum Cladding Oxidation

< 17% 7.20%Maximum Core-Wide Oxidation

< 1% < 0.60%Coolable Geometry Yes Yes As the bounding SBLOCA analysis is found to comply with all SER limitations and conditions, and all acceptance criteria for 10 CFR 50.46 are met, the associated bounding LBLOCA AOR bounds operation at the proposed MUR power uprate of 2737 MWt plus uncertainties.

28 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION 11.4 ANTICIPATED TRANSIENTS WITHOUT SCRAM As noted in Reference 11-60, Calvert Cliffs has installed aDiverse Scram System. The NRC concluded that the Diverse Scram System met the requirements of 10 CFR 50.62 in Reference 11-61.Reference 11-62 stated that the installation of the Diverse Scram System, diverse turbine trip, and diverse Auxiliary Feedwater Actuation System (AFAS), maintain the probability and consequences of an Anticipated Transients Without Scram (ATWS) as low, and eliminate the need to consider an ATWS as a DBE. Therefore, the proposed MUR power uprate does not adversely impact ATWS.11.5 CONTAINMENT RESPONSE The mass and energy transfer data for the limiting LOCA DBA is based on three types of LOCA DBAs;hot leg LOCA with minimum SI, cold leg LOCA with minimum SI, and cold leg LOCA with maximum SI. The limiting LOCA DBA is the cold leg LOCA with maximum SI. The limiting LOCA DBA assumes an initial reactor power of 102% (2754 MWt).The mass and energy for the Main Steam Line Break (MSLB) DBA includes.

a spectrum of core power levels to determine the most limiting mass and energy transfer for containment peak pressure and temperature including 0%, 50%, 75%, and 102% power levels. The most limiting for MSLB DBA corresponds to a 75% power level.Note that all other events that challenge the containment integrity and are mentioned in other UFSAR sections are bounded by the limiting LOCA and MSLB DBA analyzed in Section 14.20 and discussed above.11.5.1 Containment Response (UFSAR 14.20)The Containment Response is a DBE, the analysis of which verifies the integrity of the containment structure under the adverse pressure and temperature conditions resulting from a postulated LOCA or MSLB Event. Parametric combinations of break size, break location, and power level are analyzed to determine the most limiting scenario with specific regard to containment response for both LOCA and MSLB. Design and acceptance criteria are placed on the limiting temperature and pressure results, which ensure the integrity of the containment structure under the conditions of the analyzed events.Reference 11-31 is the current bounding AOR for containment response, applicable to plant conditions with and without the replacement SGs, and valid beyond a rated power level of 2737 MWt (MIUR power uprate), In support of the replacement SG installation, the bounding AOR (Reference 11-63) was established.

Reference 11-64 provides the qualification of the GOTHIC computer code for modeling containment response at Calvert Cliffs. This methodology was implemented at Calvert Cliffs in accordance with the 10 CFR 50.59 process, as documented in Reference 11-65. Limiting mass and energy releases are determined parametrically, and include power levels of 2754 MWt. Decay heat values following the modeled plant trip are calculated based on the NRC Branch Technical Position ASB 9-2 for LOCA. The MSLB results bound those of LOCA with respect to both peak pressure and peak temperature in Containment during the course of the analyzed limiting events. The limiting initial power level for the MSLB event is 75% RTP, however a power level of 2754 MWt, plus pump heat, was analyzed parametrically with various break sizes to determine the limiting contribution of mass and energy to the containment atmosphere through the break. All documented bounding results in Reference 11-31 (AOR) bound operation at the MUR power uprate power level of 2737 MWt, plus uncertainties, and are found to be in compliance with the applicable qualification restraints of Reference 11-64. Therefore, the current AOR for the Containment Response Analysis is appropriately applicable to, and bounds, operation at the MUR power uprate power level of 2737 MWt plus uncertainties.

29 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION 11.6 STATION BLACKOUT EVENT The proposed changes to the licensed core power level and power measurement uncertainty have no impact on the station blackout analysis.

The initial portion of the station blackout transient (i.e., loss of AC power) was determined to be unaffected by the proposed MUR power uprate (see Table I1-1). The small increase in decay heat as a result of the proposed MUR power uprate has a negligible impact on post-trip equipment (e.g., opening of MSSVs) or operator response.

11.7 REFERENCES

11-1 NRC Regulatory Issue Summary 2002-03, Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications, dated January 31, 2002 11-2 Letter from A.W. Dromerick (NRC) to C.H. Cruse (BGE), dated May 23, 1998, Docket Nos.50-317 and 50-318, "Issuance of Amendments for Calvert Cliffs Nuclear Power Plant Unit No. 1 (TAC No. M97855) and Unit No. 2 (TAC No. M97856)" 11-3 SE00495, Revision 0003, "Unit 2 Cycle 16 Core Reload (2005 RFO)," March 11, 2005 11-4 CA06386, Revision 0001, "Calvert Cliffs Units I & 2 Control Element Assembly Withdrawal Event," December 14, 2004 11-5 Letter from D.H. Jaffe (Signed by R.A. Clark) (NRC) to A.E. Lundvall, Jr. (BG&E), dated June 24, 1982, Amendment No. 71 to Facility Operating License No. DPR-53 for Calvert Cliffs Nuclear Power Plant, Unit No. 1 Letter from D.H. Jaffe (NRC) to A.E. Lundvall, Jr. (BG&E), dated January 10, 1983, Amendment No. 61 to Facility Operating License No. DPR-69 for Calvert Cliffs Nuclear Power Plant, Unit No. 2 11-6 CA06389, Revision 0000, "Calvert Cliffs Units I & 2 Excess Load Event," April 13, 2004 11-7 SE00471, Revision 0001, "Unit 1 Cycle 16 Reload Physics and Transients Safety Evaluation," November 21, 2002 11-8 CA05745, Revision 0000, "Calvert Cliffs Units I & 2 Loss of Load Transient Analysis," February 1, 2002 11-9 Letter from D.M. Skay (NRC) to C.H. Cruse (CCNPP), dated February 26, 2002, Docket Nos.50-317 and 50-318, "Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 -Amendment RE: Reanalysis of Loss of Feedwater Event (TAC Nos. MB3442 and MB3443)" 11-10 CA05733, Revision 0001, "Calvert Cliffs Units I & 2 Loss of Feedwater Flow Event," February 22, 2002 11-11 CA03552-0001, "Revision Completed to Support UIC16 Reload Including Replacement Steam Generators," February 22, 2002. [Calvert Cliffs Owner Acceptance Review of Westinghouse Calculation A-CC-FE-0029, Revision 02, "Calvert Cliffs Units 1 and 2 RCS Depressurization Event Analysis"]

11-12 CA06509, Revision 0000, "Calvert Cliffs Units 1 & 2 Loss of Coolant Flow Event," February 23, 2005 11-13 CA3553-001, Calculation Change Notice for "BGE Calvert Cliffs Units I and 2 Loss of Non-Emergency AC Power Evaluation for Reduced Flow and 2500 Plugged Tubes," Addressing Revision Completed to Support U1C16 Reload Including Replacement Steam Generators, February 22, 2002 30 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION 11-14 CA06385, Revision 0000, "Calvert Cliffs Units 1 & 2 Control Element Assembly Drop Event," March 30, 2004 11-15 SE00492, Revision 0000, "Unit 1 Cycle 17 Reload Physics and Transients Safety Evaluation," April 23, 2004 11-16 CA06508, Revision 0001, "Calvert Cliffs Units 1 & 2 Control Element Assembly Ejection Event," March 5, 2006 11-17 SE00498, Revision 0001, "Unit 2 Cycle 17 Reload Physics and Transients Safety Evaluation," March 26, 2007 11-18 SE00499, Revision 0003, "Unit I Cycle 19 Reload Physics and Transients Safety Evaluation," March 29, 2008 11-19 CA06383, Revision 0000, "Calvert Cliffs Units I & 2 Pre-Trip Steam Line Break Event," March 15, 2004 11-20 CA06917, Revision 0000, "Calvert Cliffs Unit I Cycle 19 Post-Trip Steam Line Break Event," February 29, 2008 11-21 CA06790, Revision 0000, "Calvert Cliffs Unit 2 Cycle 17 Post-Trip Steam Line Break Event," March 23, 2007 11-22 A-CC-FE-0067, Revision 07, "Calvert Cliffs SGTR Event with EOP-Based Operator Actions and Isolated ADVs," December 15, 2003 11-23 CA06384, Revision 0000, "Calvert Cliffs Units I & 2 Seized Rotor Event," March 30, 2004 11-24 Letter from D.G. McDonald (NRC) to G.C. Creel (BGE), "Issuance of Amendment for Calvert Cliffs Nuclear Power Plant Unit No. 1 (TAC No. M82277)," May 26, 1992 11-25 CA06551, Revision 0001, "Calvert Cliffs Units 1 and 2 SBLOCA ECCS Performance Analysis for Implementation of ZrB 2 IFBA and Axial Blankets," February 21, 2006 11-26 CA06550, Revision 0000, "Calvert Cliffs Units 1 and 2 1999 EM LBLOCA ECCS Performance Analysis for Implementation of ZrB 2/Axial Blankets," January 15, 2005 11-27 NEU 94-030, Revision 0, "Offsite Doses at the Exclusion Area Boundary with a Fuel Handling Incident in the Spent Fuel Pool Area," December 22, 1994 11-28 000-DA-9302, Revision 1, "Re-evaluation of Fuel Handling Accident Supporting Both Personnel Air Lock Doors Open During Fuel Movement (Open Door Policy)," October 13, 1993 11-29 C.E. Rossi (NRC) to J.A. Martin (Westinghouse Electric Corporation), "Safety Evaluation Report, dated February 2, 1987, Approval for Referencing of Licensing Topical Reports: March 1974 Report; WSTG-2-P, May 1981; and WSTG-3-P," July 1984 11-30 Turbine Missile Analysis Statement, Constellation Nuclear, Calvert Cliffs Unit 1, TB.# 170X413, Mark S. Page, GE Energy. Services, July 3, 2003 11-31 CA06774, Revision 0000, "Containment Response to LOCA and MSLB for Calvert Cliffs Units 1 and 2," March 26, 2007 11-32 CA05994, Revision 0000, "RC Waste Processing System Incident and Waste Gas Incident Dose Analysis," October 18, 2002 11-33 M-89-583, Revision 3, "Offsite and Control Room Doses Following a LOCA," July 9, 1991 11-34 CA03746, Revision 0000, "Excess Charging Event," August 28, 1998 31 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION 11-35 CA0621 1, Revision 0000, "Calvert Cliffs Units I & 2 Feedwater Line Break Event," March 7, 2003 11-36 S.A. Richards (NRC) to I.C. Rickard (ABB-CE), "Acceptance for Referencing of CENPD-387-P, Revision-00-P, 'ABB Critical Heat Flux Correlations for PWR Fuel' (TAC NO. MA6109)," March 16, 2000 11-37 K. Kniel (NRC) to A.E. Scherer (CE), "Evaluation of Topical Report CENPD-161-P

["TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core"]," September 14, 1976 11-38 CEN-191(B)-P, "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units I and 2," December 1981 [CETOP-D]11-39 C.O. Thomas (NRC) to A.E. Scherer (CE), "Combustion Engineering Thermal-Hydraulic Computer Program CESEC-IlI," April 3, 1984 11-40 Letter from S.A McNeil, Jr. (NRC) to J.A. Tiernan (BGE), "Safety Evaluation of Topical Report CEN-348(B)-P, 'Extended Statistical Combination of Uncertainties,"'

October 21, 1987 11-41 CA03552, Revision 00, "Calvert Cliffs Units 1 and 2 RCS Depressurization Event Analysis for the Low Flow Reduction Project," January 20, 1997 11-42 H.N. Berkow (NRC) to J.A. Gresham (WEC), "Final Safety Evaluation for Topical Report ,WCAP-16072-P, Revision 00, 'Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs,"'

May 6, 2004 11-43 CENPD-183-A (incl. Amendment 1-P), "Loss of Flow -C-E Methods for Loss of Flow Analysis," June 1984 H. Bernard (NRC) to A.E. Scherer (CE), "Acceptance for Referencing of Licensing Topical Report CENPD-183," May 12, 1976 11-44 CENPD-188-A, "HERMITE:

A Multi-Dimensional Space-Time Kinetics Code for PWR Transients," March 1976 O.D. Parr (NRC) to A.E. Scherer (CE), June 10, 1976 11-45 CA06388, Revision 0000, "Calvert Cliffs Unit 1 & 2 Asymmetric Steam Generator Event," March 15, 2004 11-46 LTR-ESI-01-224, "Limitation/Constraint Identification in the NRC SER for CENPD-404-P, Rev. 0, 'Implementation of ZIRLOTM Cladding Material in CE Nuclear Power Fuel Assembly Designs,'" C. M. Molnar, December 7, 2001 11-47 CENPD-190-A, "CEA Ejection, C-E Method for Control Element Assembly Ejection," July 1976 O.D. Parr (NRC) to A.E. Scherer (CE), June 10, 1976 11-48 R.L. Baer (NRC) to A.E. Scherer (CE), "Evaluation of Topical Report CENPD-135 Supplement 5," September 6, 1978 11-49 Letter from R.C. Clark (NRC) to A.E. Lundvall, Jr. (BGE), No Title, July 15, 1983 11-50 C.O. Thomas (NRC) to A.E. Scherer (CE), "Acceptance for Reference of Licensing Topical Report CENPD-207(P/NP), C-E Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids, Non-Uniform Axial Power Distribution," November 2, 1984 11-51 A-CC2-FE-0097, Revision 003, "Calvert Cliffs Unit 2 Cycle 14: Evaluation of Non-LOCA Transient Analyses and Summary of Set point Analysis Inputs," January 12, 2001 32 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION 11-52 LTR-TAS-04-100, Revision 0, "Assessment of Scram Curve Changes on Various Non-LOCA Safety Analyses for, Calvert Cliffs Unit 2 Cycle 16," September 21, 2004 11-53 CENPD-132, Supplement 4-P-A, "Calculative Methods for the CE Nuclear Power Large Break LOCA Evaluation Model," March 2001 11-54 CENPD-137, Supplement 2-P-A, "Calculative Methods for the ABB CE Small Break LOCA Evaluation Model," April 1998 11-55 Letter from S.A. Richards (NRC) to P.W. Richardson (Westinghouse), dated December 15, 2000,"Safety Evaluation of Topical Report CENPD-132, Supplement 4, Revision 1, 'Calculative Methods for the CE Nuclear Power Large Break LOCA Evaluation Model' (TAC No. MA5660)" 11-56 Letter from T.H. Essig (NRC) to I.C. Rickard (ABB CENP), dated December 16, 1997,"Acceptance for Referencing of the Topical Report CENPD-137(P), Supplement 2, 'Calculative Methods for the C-E Small Break LOCA Evaluation Model' (TAC No. M95687)" 11-57 Letter from D.M. Skay (NRC) to C.H. Cruse (CCNPP), dated April 8, 2002, "Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 -Amendment RE: Implementation of ZIRLO Clad Fuel Rods (TAC Nos. MB2540 and MB2541)" 11-58 Letter from C.H. Cruse (CCNPP) to Document Control Desk (NRC), dated May 9, 2002,"Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2, Docket Nos. 50-317 & 50-318, 10 CFR 50.46 30-Day Report for Changes to the Calvert Cliffs Nuclear Power Plant Emergency Core Cooling System Performance Analysis" 11-59 CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model," August 1974 11-60 Letter from A.W. Dromerick (NRC) to C.H. Cruse (BGE), dated October 2, 1997, "Issuance of Amendments for Calvert Cliffs Nuclear Power Plant Unit 1 (TAC No. M95181) and Unit No. 2 (TAC No. M95182)" 11-61 Letter from S.A. McNeil (NRC) to J.A. Tiernan (BGE), dated November 2, 1988, "Safety Evaluation Concerning Conformance to the ATWS Rule (TACs 59079 and 59080)" 11-62 Letter from C.H. Cruse (BGE) to Document Control Desk (NRC), dated July 31, 1997,"Response to Request for Additional Information Regarding the Technical Specification Change to the Moderator Temperature Coefficient (TAC Nos. M95181 and M95182)" 11-63 CA05892, Revision 0001, "Containment Response to OSG and RSG DBA for USAR," May 15, 2002 11-64 CA03559, Revision 0000, "Topical Report, GOTHIC Code Containment Response Analysis Model Qualification," April 9, 1998 11-65 SE00040, 10 CFR 50.59 Safety Evaluation, "Updated FSAR 14.20 Containment Response Safety Evaluation," December 20, 1995 11-66 Letter from D.V. Pickett (NRC) to J.A. Spina (CCNPP), dated August 29, 2007, Calvert Cliffs Nuclear Power Plant, Unit Nos. I and 2 -Amendment RE: Implementation of Alternative Radiological Source Term (TAC Nos. MC8845 and MC8846)11-67 Letter from B.S. Montgomery (CCNPP) to Document Control Desk (NRC), dated November 3, 2005, License Amendment Request: Revision to Accident Source Term and Associated Technical Specifications 11-68 CA06450, Revision 0000, "Fuel Handling Accident Using Alternate Source Terms,", July 1, 2005 33 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION III. ACCIDENTS AND TRANSIENTS FOR WHICH THE EXISTING ANALYSES OF RECORD DO NOT BOUND PLANT OPERATION AT THE PROPOSED INCREASED POWER LEVEL There are no accidents or transients that are not bounded by the existing AOR (see Table I-1). However, other related personnel and equipment concerns need to be addressed.

Therefore, the potential effects of the MUR power uprate were evaluated for the following issues: " Normal Operational Shielding and Personnel Exposure" Radiological Environmental Qualification (EQ)" Post-LOCA Access to Vital Areas As discussed in the previous section, no Chapter 14 accidents or transients required additional analysis because the existing AOR remained bounding for plant operation at the proposed increased power level.Discussion on the impact of the proposed MUR power uprate on plant radioactive waste effluents is provided in Section VI under Radioactive Waste Systems.Normal Operational Shielding and Personnel Exposure The MUR power uprate is expected to cause a 1.38% increase in radiation levels. However, these increases will not affect radiation zoning or shielding requirements in the various areas of the plant.Individual worker exposures are maintained within acceptable limits by the site as low as reasonably achievable (ALARA) program that controls access to radiation areas. In addition, procedural controls may be used to compensate for increased radiation levels.Radiological Environmental Qualification In accordance with 10 CFR 50.49, safety-related electrical equipment must be qualified to survive the radiation environment at their specific location during normal operation and during an accident.The Containment and Auxiliary Buildings are divided into various rooms for environmental zoning purposes.

The radiological environmental conditions noted for these rooms are the maximum conditions expected to occur. The current normal operation values represent 40 years of operation, while the AOR post-accident radiation exposure levels are determined for a one-year period following an accident using Regulatory Guide 1.89 source-term assumptions and a core power level of 2700 MWt.For the MUR power uprate, the EQ accident source-term was reanalyzed for a core power level that bounds the proposed MUR power uprate with the same release assumptions as before. The increased source-term was compared to the AOR to develop integrated energy ratios that were used to adjust the doses from various sources (airborne, sump, iodine filters, etc.) for each Containment and Auxiliary Building room. The normal operation contribution to the EQ dose is based on survey data. It was increased by 1.38% (M1UR power uprate), as well as by a factor of 1.5 to account for the extended operation period of 60 years.Post-LOCA Access to Vital Areas Vital access dose considerations are described in NUREG-0737, Item II.B.2. Specifically, the design dose for personnel in a vital area should not exceed 5 rem whole body, or its equivalent to any part of the body, for the duration of DBAs. Updates of the dose analyses were performed to confirm that this requirement was met for a LOCA using Regulatory Guide 1.4 source-term assumptions and a core power level of 2737 MWt. The UFSAR time-dependent radiation dose rate maps that cover plant areas and access paths which may require occupancy during post-LOCA recovery operations will be updated to 34 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION reflect the proposed MUR power uprate. The MUR power uprate does not have an impact on vital area access requirements.

35 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION IV. MECHANICAL/STRUCTURAL/MATERIAL COMPONENT INTEGRITY AND IV.A INTRODUCTION The RCS component specifications define the frequency and severity of the design transients that must be considered in the fatigue evaluations of the components in accordance with the American Society of Mechanical Engineers (ASME) code. The design transients in the individual component specifications represent events that are expected to occur, or may occur, during the life of the plant. The design transients are characterized in terms of the type of transients, the frequency of occurrence, the initial design conditions, and the associated thermal-hydraulic conditions experienced by various systems and components as a result of the transients.

This information is then used in fatigue evaluations for those systems and components.

The design transients defined in the current component specifications were reviewed to determine the effect of the MUR power uprate.With respect to the type of transients and frequency of occurrence, the implementation of the MUR power uprate does not create neN. types of transients nor change the original event frequencies for the design transients, With respect to the initial conditions and the thermal-hydraulic response during the transients, some were found to be affected by the uprate and some were not. The transients which occur in the lower operating modes remain valid because the HZP (no load) plant conditions are unaffected by the MUR power uprate.Many of the transient responses remain valid because the original design hot and cold leg temperatures are higher than the increased operating point due to the MUR power uprate.Where necessary, the design transients were re-analyzed quantitatively to assess the impact of the changes on existing design conditions due to the MUR power uprate. In these cases, the analyses simulated the transients under the increased conditions and produced thermal-hydraulic responses (pressures, temperatures, and flow rates) for use in the component-by-component fatigue evaluations described in this section.This section also provides the results of the structural integrity evaluations for RCS components and supports at MUR power uprate conditions.

Table IV-1 shows a comparison of current nominal operating parameters values versus the expected values following implementation of the MUR power uprate. The remaining portions of this section discuss the impact of the MUR power uprate RCS components, nuclear fuel and core thermal hydraulics, and various other components.

36 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION Table IV-1 Current NSSS Design and MUR Power Uprate Nominal Operating Parameters for Calvert Cliffs Parameter Core Power, MWt (input)No. of Plugged Tubes per SG Primary Bulk Th, IF Primary T,, IF Primary AT, IF Primary Flow Rate, gpm (input)Core Bypass Flow Rate, %Primary Pressure, psia Feedwater Temperature, IF Feedwater Enthalpy, Btu/lbm (input)Feedwater Flow Rate per SG, Ibm/sec (input)SG Blowdown Flow per SG, Ibm/sec (input)SG Steam Flow per SG, Mlbm/hr Steam Pressure, psia Current Normal Operating Conditions 2700< 10%595.1 548 48.4 (3)370,000 -422,250 3.9 2250 431.5 (5)409.2(,5)Same as Steam Flow 41.7 (max)(6)MUR Power Uprate Normal Operating Conditions 2746 (6)< 10%595.9(7)548 49.2 (3)370,000 -422,250 3.9 2250 433.6 (5 410.8 (1,5)Same as-Steam Flow 41.7 (max)(6)Percent Change 1.70% (6)0.13%1.65%0.70%0.39%5.9000)888 (1,2)863 (1,3)5.999 (1, 5,6)886.5 ("'2)860.3 (1,3)1.68%-0.17%-0.31%SG Total Mass, Ibm 138,524 (,,4) 138,024 (, 4) -0.36%SG Liquid Mass (Ibm) 128,130 (1,4) 127,636 (1,4) -0.39%() At 100% power (2) No plugged tubes (3) 10% plugged tubes (4) SG level at 35.95 ft (5) Based on best available data (6) Bounding value selected for the evaluation (7) A large portion of the MUR power uprate evaluation was completed using an estimated temperature increase for Thot of 1. I°F. Further evaluations have since been finalized, predicting a 0.8'F increase for Thot. Therefore, the original evaluation performed for MUR power uprate remains bounding.IV.2 REACTOR COOLANT SYSTEM LOSS-OF-COOLANT ACCIDENT FORCES EVALUATION The purpose of a LOCA hydraulic forces analysis is to generate the hydraulic forcing functions and blowdown loads that occur on RCS components as a result of a postulated LOCA. These forcing functions and loads act on the component's shell and internal structures.

The full set of RCS loadings considered in the structural analysis of a LOCA event consists of the internal forcing functions generated from the hydraulic forces analysis, the pipe tension release, and jet impingement forces acting at the break locations, and, where applicable, the external loads due to subcompartment pressurization effects that act on the components and their supports.Except for the thimble support plate and selected .RV internals components, the faulted loads and stresses in the current AORs are based on main coolant loop breaks (MCLBs) where thrust loadings were based on simplified (pressure x area) terms and where asymmetric blowdown loadings were calculated using 37 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION design setpoint parameters for 2700 MWt, where Thot=604°F and TCOId=548°F.

Since the MCLBs have been eliminated by leak before break (LBB) and replaced by branch line pipe breaks (BLPBs), loads and motions on NSSS components due to pipe breaks are greatly reduced. Furthermore, since the RV blowdown loads are primarily affected by changes in TCOid, and Tcold remains the same for the MIUR power uprate, the effects of BLPBs at the MUR power uprate would not be significantly different from the effects of BLPBs under pre-MUR power uprate conditions.

Therefore, the effects on NSSS components of BLPBs at the MUR power uprate are less severe than the effects of pipe breaks currently documented in the AORs.Based on this conclusion, the design transient for blowdown loads at the MUR power uprate conditions remains the original design basis LOCA analyzed.

Except where noted, the following structural evaluation discussions are based on the original design transient, and do not make direct use of the mitigating effects of LBB.IV.3 REACTOR COOLANT SYSTEM MAJOR COMPONENT ASSESSMENTS As noted in the introduction to this section, the majority of the NSSS design transients are demonstrated to be unaffected by the MUR power uprate. Transients with the potential to adversely affect the AOR results for particular RCS components were evaluated for their effects on the critical stress margins identified for the RCS components.

The transients involved are listed below, on a component by component basis: RV, RCPs, RCS Piping and Fittings (except Surge Line), and Original Control Rod Drive Mechanism

&Part Length Control Rod Drive Mechanisms Reactor Trip -The rate of change in temperature for the MUR power uprate for this transient is slightly greater than that for the design basis.Surge Line and Fittings Reactor Trip, Loss of Flow, Step Load Increase/Decrease, Plant Loading/Unloading

-The change in temperature for the MUR power uprate for these transients is slightly greater than that for the design basis.Pressurizer

• Step Load Increase -The rate of change in temperature for the MUR power uprate is greater than that for the design basis for this transient.

The above observations were used to help determine which MUR power uprate transients needed to be evaluated with respect to their effects on fatigue for limiting RCS components.

Evaluations of these limiting components are discussed in the remainder of this section.In another assessment, the Calvert Cliffs RCS loads and displacements due to normal operating thermal expansion effects under the MUR power uprate conditions were reconciled with the loads and displacements from the pre-uprate RCS thermal expansion analysis, where Thot was set to 604'F and Tcold was set to 550'F. It was concluded that MUR power uprate does not cause any significant changes in thermal anchor motions, and that all previously documented thermal anchor motions for Calvert Cliffs remain valid. All RCS loads due to normal operating thermal expansion either decrease or change insignificantly due to the decrease in delta-T between ToId and Thot of the initial power rating design setpoint temperature and the MUR power uprate conditions stated in Table WV-1. The SG inlet nozzle 38 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION moment (the moment around the horizontal axis perpendicular to the hot leg axis) increases, but this moment is not a limiting load with respect to stress margins on either the hot leg or the SG inlet nozzle.Based on the results of this normal operating thermal expansion evaluation, specified normal operating loads on NSSS component supports and nozzles, and the main loop piping, and normal operating displacements on RCS tributary nozzles do not need to be revised for the MUR power uprate. This conclusion is utilized in the AOR stress evaluations discussed in the remainder of this section.IV.3.1 Reactor Vessel Structural Evaluation This evaluation assesses the effects that the MUR power uprate has on the most limiting locations with regard to ranges of stress intensity and fatigue usage factors (UFs) in each of the vessel regions, as identified in the RV stress reports and addenda.The nominal vessel outlet temperature increases to 595.9'F (597.2°F end-of-life), and the nominal vessel inlet temperature remains at the current value of 548.0'F as a result of the MUR power uprate (see Table IV-1 for a comparison of operating parameters).

Therefore, the Thot variation during normal plant loading and plant unloading increases while the ToId variation remains unchanged.

As noted above, the nominal vessel inlet temperature associated with the MUR power uprate is the same as the nominal temperature for the current fuel cycle. The nominal vessel outlet temperature has increased slightly but is still less than the normal design vessel outlet temperature of 604°F that was originally used in the analysis of the RV outlet nozzles. Therefore, the effects of the plant loading and unloading transients on the inlet and outlet nozzles remain bounded by the stress AOR.The RV main closure flange region and CEDM housings were originally evaluated for the effects of a higher vessel outlet temperature.

Therefore, the effects of the MUR power uprate vessel outlet temperature on these regions are also bounded by the current design basis.The remaining RV regions, including the inlet nozzles, vessel wall transition, core support guides, bottom head-to-shell juncture, and instrumentation nozzles are affected by the vessel inlet temperature, which is unchanged for the MUR power uprate. Therefore, the previously determined maximum stress intensity ranges and maximum cumulative fatigue UFs for these regions are valid.The critical stress margin at the closure head studs remains unchanged because it is based on compression due to the bolt-up procedure, which is unchanged by the MUR power uprate. The critical margins at the vessel wall at the core stabilizer lugs remain unchanged because they are based on normal operating pressure and Operating Basis Earthquake (OBE), none of which are changed by the MUR power uprate.The margins on the load capability of the RV cold leg and hot leg horizontal supports due to MCLBs are significantly increased to non-critical margins due to the elimination of these breaks and their replacement with BLPBs.None of the margins on the incore instrumentation (ICI) flange assembly for either unit are critical.

The lowest margin on stress is 7.7%, which is a margin on the bearing stress at the nut-to-compression collar surface for the flange assembly on Unit 2. This margin is unchanged because the bearing stress is due to design pressure and OBE, neither of which are changed by the MIUR power uprate.None of the margins on the RV vent line are considered to be critical.

The lowest margin on stress is for the primary-plus-secondary stress at the J-weld on the RV closure head. The controlling stress range is generated from the loss of secondary pressure (LOSP) and the normal heatup transients.

These specified 39 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION design transients are not changed by the MUR power uprate. Therefore, the critical stress margins for this component remain valid for MUR power uprate conditions.

The critical stress margins for other RV components are discussed below.RV Closure Head Instrument Nozzle Bimetallic Weld The critical margin of 10.18% on the stress intensity is due to a design pressure of 2500 psia, and the allowable Sm is based on a design temperature of 650'F. The design pressure and temperature are not changed due to the MUR power uprate. Therefore, the stress margin for this component is unchanged and remains acceptable for the MUR power uprate.Vessel Wall at RV Outlet Nozzle The critical stress margin of 2.5% is for primary-membrane-plus-local stress. Per the AOR, the calculated stress is a function of design moments and forces on the pipe, and of a design pressure of 2500 psia. The design moments and forces are unchanged due to the MUR power uprate. In addition, the design pressure of 2500 psia is unchanged by the MUR power uprate. Therefore, this stress margin remains acceptable for the MUR power uprate.RV Outlet Nozzle As above, the critical stress margin for primary-membrane-plus-local stress is a function of design moments and forces on the pipe, and of a design pressure of 2500 psia. Therefore, the stress margin of 0.79% for this component is unchanged and remains acceptable for the MUR power uprate.Vessel Wall Transition Part of Vessel Support The critical stress margin of 0.64% is for the primary-membrane stress. Per the AOR, the calculated primary-membrane stress is based on a design pressure of 2500 psia. The design pressure of 2500 psia is not changed by the MUR power uprate. Therefore, the stress margin of 0.64% for this component is unchanged and remains acceptable for the MUR power uprate.Taper between RV Dome and Closure Flange The critical stress margin of 32.0% is for primary-membrane-plus-local stress. Per the AOR, the calculated stress is due to design pressure, flange bolt-up loads and core (i.e., vessel internals) loads. The specified flange bolt-up loads and core loads are unchanged due to the MUR power uprate. In addition, the design pressure of 2500 psia is not changed by the MUR power uprate. Therefore, this stress margin is unchanged and remains acceptable for the MUR power uprate.Surveillance Holder and Brackets The critical margin on the alternating stresses (Salt) from peak stresses is 2.33%. Per the AOR, the calculated stress is based on design moments and forces and stress concentration factors. None of these parameters are changed as a result of the MUR power uprate. Therefore, the stress margin for this component is unchanged and remains acceptable for the MUR power uprate.Vessel Wall at Core Stabilizer Lugs The critical margin on the maximum stress (Smax) due to the lateral load on the shell at the lug attachment is 5.25%. Per the AOR, the calculated stress is a function of the lateral moment and the design pressure.The specified design moments and forces are unchanged due to the MUR power uprate. In addition, the 40 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION design pressure of 2500 psia is not changed by the MUR power uprate. Therefore, this stress margin is unchanged and remains acceptable for the MlUR power uprate.Vessel Shell and Bottom Head The critical margin on primary-membrane stress is 0.64%. It is noted that the critical location coincides with the vessel wall transition part of vessel support location discussed above. Therefore, the stress margin for this component is unchanged and remains acceptable for the MUR power uprate.Head Lift Rig This evaluation pertains to the currently installed head lift rigs at Calvert Cliffs Units 1 and 2. Evaluation of the planned replacement lift rig will be performed prior to its installation after the MUR power uprate.The vertical link in the head lift rig has a critical margin on stress of 2.3%. This small margin is due to tension stress during closure stud handling, which is not affected by the MIUR power uprate. Therefore, this head lift rig subcomponent remains acceptable for the MUR power uprate.The cooling duct cover plate has a critical margin on stress intensity of 0.9%, which is due to a combination of dead weight, seismic excitation, and flow loads. These flow loads are hot air flow loads across the cover plate for which the operating temperature under the MUR power uprate conditions remains lower than the design temperature.

In addition, the dead weight and seismic loads are not affected by the MUR power uprate. Therefore, this critical head lift rig subcomponent also remains acceptable.

Conclusion The RV evaluation for the MUR power uprate demonstrates that the maximum ranges of stress intensity remain within their applicable acceptance criteria, and the maximum cumulative fatigue UFs remain below the acceptance criterion of 1.0.In addition, the faulted condition stress analyses for the Calvert Cliffs RV do not change as a result of the MUR power uprate, because the seismic loads are unchanged from the AOR, and the pipe break load input remains based on the original MCLBs. Therefore, no changes in the faulted condition RV/reactor internals interface loads or other faulted condition parameters are identified as a result of the MUR power uprate.IV.3.2 Reactor Vessel Internals Evaluation The reactor internals support the fuel and control rod assemblies, experience control rod assembly dynamic loads, and transmit these and other loads (e.g., deadweight, seismic vibration) to the RV. The internals also direct flow through the fuel assemblies, provide adequate cooling to various internals structures, and support ICI. The changes in the RCS design parameters identified previously in Table IV-1 produce insignificant changes in the boundary conditions experienced by the reactor internals components.

This section describes the evaluation performed to demonstrate that the reactor internals can perform their intended design functions at the MUR power uprate conditions.

IV.3.2.1 Thermal-Hydraulic Systems Evaluations The MUR power uprate can potentially affect such parameters as reactor vessel internal (RVI) component heating rates, coolant temperature levels, and their downstream impacts. A key area in evaluation of core performance is the determination of the hydraulic behavior of coolant flow and its effect within the reactor internals system. The core bypass flows are required to ensure reactor performance and adequate 41 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION RV head cooling. The hydraulic forces are critical in the assessment of the structural integrity of the reactor internals.

The results of the thermal-hydraulic evaluations are provided below.RVI Component Temperatures The AOR on RVI component temperatures were reviewed to determine the component most affected by the MUR power uprate. The component selected from this review process is the core shroud.Component metal temperatures, and therefore, thermal stresses, are dependent on the core power level and coolant temperatures.

Calvert Cliffs core shroud metal temperatures were re-evaluated for the MUR power uprate level. The resulting core shroud component temperatures were used to calculate thermal stresses, in order to evaluate the structural margins for the shroud. The structural evaluation demonstrated that there is adequate structural margin for the core shroud for the MUR power uprate; see Section IV.3.5 RVI Component Hydraulic Loads A review of the AOR design hydraulic loads on the RVI components indicated that the current design loads are bounding for the MUR power uprate operation.

Small increases in power level, such as the MUR power uprate, have minimal impact on the design hydraulic loads.Core Bypass Flow Calculation Bypass flow is the total amount of reactor coolant flow bypassing the core region and is, therefore, not considered effective in the core heat transfer process. The design core bypass flow limit is 3.90% of the total RV flow. This value is used in the thermal margin calculations.

A lower bound value of 1.6% is used in the calculation of hydraulic loads since the higher core flow results in higher core pressure drops and, therefore, higher uplift and differential pressure (D/P) loads. The best-estimate core bypass flow is 3.51% of the RV flow.Core bypass flow is negligibly affected by the MUR power uprate. The core pressure drop will tend to increase very slightly, due to the higher power level. This will have the effect of diverting very slightly more bypass flow through the various leakage flow paths. But the margin between the best estimate and design values of core bypass flow will readily accommodate the negligible increase in core bypass flow due to the uprate.Therefore, the core bypass flow limit of 3.9% remains valid for the MUR power uprate.CEA Drop Time Analyses Calvert Cliffs Technical Specification Surveillance Requirement 3.1.4.6 requires that the drop times of all full-length CEAs from a fully withdrawn position, must be verified to be less than or equal to 3.1 seconds prior to the startup of each cycle.Control element assembly drop times are explicitly confirmed to meet the times assumed in the accident analyses.

An evaluation was performed for CE fleet plants to demonstrate continued compliance with the current Technical Specification requirements based on CE fleet's robust five finger silver tip CEA design, which has not shown failure to insert at any time in life through the end-of-life core burnup. The MUR power uprate conditions will slightly increase the power level in leading rodded fuel assemblies; however the projected burnup levels and fluences are substantially less than the values assumed in the design calculations.

The assembly burnups and fluences are confirmed on a cycle specific bases to be within the values assumed in the CEA design analysis.

In addition, the CEA drop time is measured prior to the startup of each reload cycle.42 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION Finally, the fluid density has not increased for the MUR power uprate since ToId has not changed and Thor has increased only slightly.

Therefore, CEA drop times are not adversely affected by the MUR power uprate.Based on the above, the current limiting rod drop time requirements remain valid for the MUR power uprate conditions.

CEA and ICI Cooling Assessment Cooling analyses for CEAs and incore instruments were performed for the MUR power uprate. These analyses indicate that the following design criteria are met: " No coolant bulk boiling will prevail inside the CEA and ICI guide tubes.* B 4 C and AgInCd maximum temperatures will stay within the design limits of 20007F and 1400'F, respectively.

IV.3.2.2 Mechanical Evaluations As discussed previously, the MUR power uprate conditions do not affect the current design bases for seismic and LOCA loads. Therefore, it was not necessary to re-evaluate the structural effects from seismic OBE and safe shutdown earthquake loads, or from the LOCA hydraulic and dynamic loads.Furthermore, it is noted that the LOCA hydraulic and dynamic loads would be less severe if BLPBs were analyzed instead of the original design basis MCLBs.With regard to flow and pump induced vibration, the current analysis uses a mechanical design flow that does not change for the revised design conditions (see Table IV-1). The MUR power uprate conditions alter the Thot fluid density. However, this very small change in the Thot fluid density has a negligible effect on the forces induced by flow. In addition, the MNUR power uprate results in a negligible change in Tave.Therefore, the mechanical loads are not affected by the MUR power uprate conditions.

IV.3.2.3 Structural Evaluations As described in Section IV.3.3, the normal operating hydraulic loads used in the AOR for the structural evaluation of the RVI components are bounding for the MUR power uprate. Seismic and LOCA loads on the RVI components are unaffected by the MIUR power uprate operation, and the primary stresses calculated in the AOR therefore remain applicable.

The MlUR power uprate can potentially increase thermal loadings and the resulting thermal stresses in the RVI components.

Because this MUR power uprate is relatively small (-1.38%), it was concluded that potential adverse effects on the RVI structures would be confined to the core shroud, which is more sensitive than the other RVI components to minor variations in thermal loading.To quantify these potential effects, the AOR for the calculation of thermal stresses in the core shroud was reviewed.

Because of limitations in this AOR, the applicability of a more recent analysis, performed for a similar core shroud design, was investigated.

The applicability of this analysis was confirmed, and the thermal stresses calculated therein were combined with the appropriate primary stress and evaluated against acceptance criteria.

Elevated (> 800'F) temperature effects (reflecting the core shroud maximum temperature of 885°F) were considered in the determination of these acceptance criteria, and a fatigue evaluation was performed.

Under thermal loadings that encompass the MUR power uprate, the core shroud analysis determined that the maximum primary-plus-secondary stress intensity exceeded the allowable value. Therefore, a 43 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION simplified elastic-plastic analysis with attendant re-evaluation of fatigue usage was performed in accordance with the acceptance criteria.The resulting maximum primary-plus-secondary stress intensity (excluding thermal bending stresses) was 37,927 psi, which is less than the 43,800 psi allowable.

Furthermore, the cumulative usage factor (CUF)for the MUR power uprate conditions was determined to be 0.375, which is significantly less than the 1.00 allowable.

Thus, the core shroud satisfies the acceptance criteria.Increases in core thermal power will slightly increase nuclear heating rates in the RVIs, such as lower core support plate, fuel alignment plate, and core shroud. Evaluations have been performed verifying that the existing thermal-hydraulic AOR will support the MUR power uprate. Therefore, the calculated component lifetimes will envelop the component lifetimes associated with the MUR power uprate related increases in nuclear heating.IV.3.3 Control Element Drive Mechanisms The CEDMs are mounted on top of the Calvert Cliffs reactor head. These components areaffected by the reactor coolant pressure, vessel outlet temperature, and hot leg NSSS design transients.

According to Table TV-i, the vessel outlet temperature for the MUR power uprate has increased slightly to 595.91F. This small temperature increase remains well below the design operating temperature of 604'F. Therefore, no additional assessments of the impact of thermal loads on the CEDMs and CEDM nozzles are required.

The reactor coolant operating pressure (2250 psia) for the MUR power uprate conditions remains the same as originally specified for the CEDMs so no additional assessment is required for pressure considerations.

Since all critical margins on the CEDMs are maintained for the MUR power uprate, these components remain acceptable.

IV.3.4 Nuclear Steam Supply System Piping and Pipe Whip The reactor coolant main coolant loop piping system (including primary loop piping and pipe whip restraints, and tributary piping nozzles) was assessed for the MUR power uprate effects. It was concluded that these equipment designs remain acceptable and continue to satisfy design basis requirements in accordance with applicable design basis criteria, which include the criteria associated with the original design basis mechanistic LOCA breaks, when considering the operating temperature, operating pressure, and flow rate effects resulting from the MUR power uprate conditions.

The primary piping and tributary nozzles remain within allowable stress limits in accordance with ASME Section III, 1965 Edition, up to and including the Winter 1967 Addendum [and in accordance with ASME Section 1II, 1986 Edition for components with a mechanical nozzle seal assembly (MNSA)].Reconciliation of a number of critical locations on the Calvert Cliffs Units I and 2 RCS piping and fittings under the MUR power uprate conditions is summarized below.Hot and Cold Leg Piping The critical margins on the maximum primary-local-plus-bending stress intensity at the hot leg and cold leg elbows are 5.80% and 3.26%, respectively.

The calculated stresses are based on the design moments from dead weight and seismic excitation, and the design pressure of 2500 psia. The specified design loads do not change for the MIUR power uprate. Therefore, the stress margins of 5.80% and 3.26% are unchanged and remain acceptable for the MUR power uprate.44 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION Pump/Pipe Junction at the Discharge Nozzle The critical margin on the maximum primary-bending stress at the pipe-pump discharge nozzle juncture is 4.82%. The stress is a function of the design loads (including loads due to OBE) and the design pressure of 2500 psia. The specified design moments are not changed by the MUR power uprate. In addition, the design pressure is not changed by the MUR power uprate. Therefore the stress margin of 4.82% for the pipe/pump juncture is unchanged and remains acceptable for the MUR power uprate.Pump/Pipe Junction at the Suction Nozzle The CUF at this location is 0.836 (i.e., a 16.4% margin). Most of the fatigue usage (0.833 of 0.836) is due to the seismic transients.

The rest of the fatigue usage (0.003) comes from the heatup transient.

The specified normal operating loads and design seismic loads are unaffected by the MUR power uprate.Therefore, the effect of seismic on stress and fatigue is unchanged.

The fatigue usage due to the heatup transient is also unchanged because this transient occurs at zero power and is not affected by the MUR power uprate. Therefore, this component remains acceptable for the MUR power uprate.Safety Injection Nozzle The critical margin on the maximum primary-plus-secondary stress, (Local Primary Membrane Stress (PL) plus Primary Bending Stress (PB) Secondary Stress (Q)) or the primary plus secondary stress intensity (PL+PB+Q), is 2.24%, and is due to the seismic load and heatup transient and the cooldown transient when combined with the effects of design pressure, dead weight, and seismic loads. The design pressure, dead weight, and seismic loads are not affected by the MUR power uprate. The heatup/cooldown transients occur at zero power and therefore are not affected by the MUR power uprate.Therefore, this critical margin on stress is unchanged for the MUR power uprate.The highest CUF in the SI nozzle is only 0.1892, and is primarily due to an alternating stress from combinations of plant cooldown/seismic and plant heatup/cooldown transients.

The plant heatup/cooldown transients and seismic excitations are not affected by the MUR power uprate.Therefore, the CUF for the SI nozzle is unchanged, and the nozzle remains acceptable for the MUR power uprate.Hot Leg RTD and Pressure Differential Transmitter Nozzles with a MNSA The critical margin on the maximum primary-plus-secondary stress, PL+PB+Q, is 1.68% for either the hot leg RTD or pressure differential transmitter nozzle with a MNSA installed.

This limiting stress is based on the hydrostatic test and LOSP transients.

These transients are not altered by the MUR power uprate. Therefore, this component remains acceptable for the MUR power uprate.Hot Leg Drain Nozzle The critical margin on primary-local-plus-bending stress is 1.25%. The calculated stress range is based on design moments from design pressure, dead weight, thermal and design seismic effects. These effects are changed by the MUR power uprate. Therefore the stress margin of 1.25% for this nozzle is unchanged and the nozzle remains acceptable for the MUR power uprate.In conclusion, the Calvert Cliffs Units 1 and 2 primary piping and tributary nozzles remain within allowable stress limits in accordance with ASME Section III, 1965 Edition, up to and including the Winter 1967 Addendum (and in accordance with ASME Section III, 1986 Edition for components with MNSA). Furthermore, no piping or pipe restraint modifications are required as a result of the increased 45 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION power level, because conservatively determined LOCA loads due to MCLBs were used to design the pipe restraint systems.IV.3.5 Steam Generators The thermal-hydraulic performance of the SGs was analyzed for the MUR power uprate conditions.

Given the new RCS input of 595.9°F for Thor, with flow, and Tcold remaining the same as at 2700 MWt, the secondary side of the SG experiences a 1.7% flow rate increase with pressure and temperature decreasing 1.5 psi and 0.1 0 F. There is also a slight 0.36% decrease in SG inventory at 100% power. The new conditions were checked against the design, test and Level A, B, C and D stress levels specified in the ASME Code and found to be acceptable.

The internals and flow induced vibration effects were found to be negligible.

Structurally there is negligible effect. Therefore, the SGs remain fully qualified to operate at the MUR power uprate.Steam Generator Upper and Lower Supports Structural Integrity The Calvert Cliffs SG support system consists of the following components at each SG:* Lower SG supports -a sliding base, with four vertical pad supports and two lower keys.* Upper SG supports -two upper shear key supports and eight directing-acting hydraulic snubbers.Even though the operating setpoint temperatures (Thot and Tcold) for the MUR power uprate conditions are enveloped by the design setpoint temperatures used in the original design basis structural analyses of the RCS, an assessment was performed to determine the effects of the MIUR power uprate condition operating temperatures on the RCS components and supports.

This analysis concluded that the loads on the RCS supports, including the supports on the SGs, either decreased or changed insignificantly due to the decrease in delta-T between Tcold and Thot, relative to the original design basis analyses.

Therefore, the effect of RCS thermal expansion on SG support loads due to the MUR power uprate is insignificant.

Since the original seismic and LOCA loads are also unchanged for the MUR power uprate, the SG upper, and lower supports continue to be acceptable under the MUR power uprate conditions.

IV.3.6 Reactor Coolant Pumps and Motors IV.3.6.1 RCP Structural Analysis The four RCPs are installed in the cold legs of the reactor coolant loops. The RCPs are affected by the reactor coolant pressure, SG outlet temperature, and primary side cold leg NSSS design transients.

The SG outlet temperature affects both the thermal expansion and thermal transient loads on the RCPs.The nominal SG primary outlet temperature for the MUR power uprate (i.e., Tcold = 548.0°F) is the same as the current nominal and design basis temperature for the SG outlet, RCP suction and discharge and RV inlet. Consequently, RCP thermal expansion loadings for the MUR power uprate are bounded by the design condition.

The RCP supports are designed to carry loads due to normal operating conditions and seismic excitations, neither of which is changed by the MUR power uprate. Under LBLOCA, the RCP nozzles were shown to be capable of carrying the faulted condition loads that the RCP supports were not designed to carry.When LBLOCAs are replaced by BLPBs via LBB considerations, the faulted loads on the RCP supports and nozzles are significantly reduced, and the margin on the nozzle loads is increased.

46 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION The AOR also identified a critical margin on the horizontal strut load due to OBE. Since seismic excitations are unaffected by the MUR power uprate, this margin remains unchanged and acceptable.

In addition to the support system, other critical RCP components stress margins were addressed in the MUR power uprate assessments, as follows.Casing Diffuser Vanes Design Conditions

-Critical margins exist for the primary-membrane and the primary-membrane-plus-local stress intensities at Vane 8. These margins are 2.74% and 4.8 1%, respectively.

Per the AOR, the calculated membrane stress intensity is the average stress in the vane, and the calculated membrane plus local stress intensity represents the largest surface stress intensity, adjusted by removing the discontinuity bending stress. These stresses are a function of design moments and forces on the structure and the design pressure of 2500 psia. The specified design moments and forces are unchanged due to the MJUR power uprate. In addition, the design pressure of 2500 psia is unchanged by the MUR power uprate.Therefore, these diffuser vane stress margins are unchanged and remain acceptable for the MUR power uprate.Suction Nozzle Design Conditions

-The critical margin on the primary-membrane stress of the suction nozzle due to the design conditions is 2.0% and is only due to the design pressure 2500 psia. The design pressure, and therefore, primary-membrane stress, is not affected by the MUR power uprate.Emergency Conditions

-The critical margin in the suction nozzle, due to emergency conditions, involve primary-local membrane stresses.

In the AOR, the overall emergency condition stresses exceeded the ASME code primary-general stress limit of 1.2 S,,. However the 1.2 Sm limit does not include local stress effects. The overall emergency condition stresses are, however, bounded by the primary-local stress ASME code limit of 1.8 Sm which does include local stress effects. The AOR concluded that the conservatism inherent in the more restrictive primary-general membrane stress allowable was unwarranted and the primary-local stress limit of 1.8 Sm was an acceptable bound for the suction nozzle stresses.

This reasoning also is applicable to the stresses for the MUR power uprate.Furthermore, since the specified external moments and forces, and the operating pressure loads are unchanged for the MUR power uprate conditions, the stress margins are unaffected and the suction nozzle design remains acceptable.

Discharge Nozzle Design Conditions

-Critical margins exist for the primary-membrane and the primary-membrane-plus-bending stress intensities in the crotch region of the discharge nozzle. These margins are 2.20% and 2.67%, respectively.

Per the AOR, the acceptable primary-membrane stress margin of 2.20% was obtained after correcting the as-calculated stress analysis results for the as-cast thickness of the discharge nozzle and shell. Regarding the primary-membrane-plus-bending stress, the acceptable margin of 2.67%was obtained by removing the secondary bending stress from the greatest surface stress intensity in the crotch. The specified design moments and forces are unchanged due to the MUR power uprate. In addition, the design pressure of 2500 psia is unchanged by the MUR power uprate. Therefore, the stress margins of 2.20% and 2.67% for the discharge nozzle remain acceptable for the MUR power uprate.Emergency Conditions

-Critical margins exist for the primary-membrane and the primary-local-plus-bending stress intensities in the crotch region of the discharge nozzle. Primary-local-plus-bending results apply to the top half of the nozzle. These margins are 0.10% and 0.87%, respectively.

Since the external 47 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION moments and forces, and the operating pressure loads are unchanged for the MUR 'power uprate conditions, the stress margins remain acceptable for the MUR power uprate.Hanger Bracket Design Conditions

-The critical margin for this location, 2.67%, is associated with the primary-membrane-plus-bending stress intensity.

The membrane stress, which is 1.046 Sm, is classified as primary-local stress, and therefore is well below the limit of 1.5 Smn. The specified design moments and forces are unchanged due to the MUR power uprate. In addition, the design pressure of 2500 psia is unchanged by the MUR power uprate. Therefore the stress margin of 2.67% for the hanger bracket remains acceptable for the MUR power uprate.Volute, Lower Flange Design Conditions

-The critical margin for this location, 4.70%, is associated with the primary-membrane stress intensity.

The AOR also states that no surface stress exceeds the 1.5 Sm limit.Consequently, the primary-membrane-plus-bending limits are also satisfied for this region of the structure.

The specified design moments and forces are unchanged due to the M1UR power uprate. In addition, the design pressure of 2500 psia is unchanged by the MUR power uprate. Therefore, the stress margin of 4.70% for the volute/lower flange region remains acceptable for the MUR power uprate.Cover, Region 4 Design Condition

-According to the AOR, the critical stress margin for this region of the cover (the inside comer of the cover between cooling holes) occurs under operating conditions (i.e., for operation between steady-state hot and steady-state cold conditions).

In this case, the highest stress intensity range was determined from linearized surface stresses and compared to 3 Sm at operating temperature, resulting in the critical margin of 4.96%. Since the heatup and cooldown transients are not affected for the MUR power uprate, the critical margin of 4.96% is unchanged for the MUR power uprate conditions.

Based on the above discussions, it can be concluded that the existing RCP stress analyses are bounding and remain applicable for the pressure boundary components.

IV.3.6.2 RCP Motor Evaluation Previous analyses determined that the RCP motors are acceptable for continuous operation with limiting hot loop and cold loop conditions under 2700 MWt. The RCP motors were determined to remain acceptable for operation at the MUR power uprate parameters based on the following:

• No-load Tave is unchanged by the MUR power uprate. Therefore, the RCP hot start is not affected.* Limiting RCP motor starting conditions occur during RCS cold loop conditions that are unchanged, and therefore not impacted by the MUR power uprate (i.e., ToId remains at the design value of 548°F)." The mechanical loads controlling RCP motor thrust bearing design are associated with seismic and LOCA conditions (i.e., RCP motor peak accelerations).

Seismic loads are not affected by the MUR power uprate, and LOCA condition loads are reduced when BLPBs are invoked as the limiting design basis pipe breaks.* The thermal transients on the drive motor are not affected by the MUR power uprate.48 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION IV.3.7 Pressurizer IV.3.7.1 Pressurizer Vessel The conditions that could affect the primary-plus-secondary stresses, and the primary-plus-secondary-plus-peak stresses, are the changes in the RCS hot leg temperature (Thot), the RCS cold leg temperature (Tc.1d), and the pressurizer transients.

Table IV-1 indicates that Twod is unchanged, and that the increase in Thor is very small. A Thor change of this magnitude is enveloped by the current stress analysis.

Some of the calculated thermal transients, however, were affected by the MUR power uprate. Therefore, critical locations in the pressurizer were re-examined, as discussed below.Pressurizer Upper, Bottom and Side RTD Nozzles, and Heater Sleeves The maximum CUF for these pressurizer locations, after NMNSA repairs, is 0.863. In all cases, the CUFs were entirely due to fatigue usage from plant heatup/cooldown and leak test transients.

None of these transients are affected by the MUR power uprate; therefore the CUFs for these components are unchanged, and the components remain acceptable for the MUR power uprate.Surge Nozzle at the Pressurizer End The surge nozzle at the pressurizer end has a CUF of 0.764. Per the AOR, a UF of 0.716 (or greater than 94% of the CUF) is due to contributions from the normal plant variations at steady-state transient and from the step load increase transient.

The normal plant variations at steady-state transient (defined as +/-100 psi and +/-67F) is unchanged by the MUR power uprate. The effect of the MUR power uprate on the step load increase transient was evaluated by calculating stress factors based on a comparison of the calculated transient based on the MUR power uprate setpoints vs. the originally specified transient.

The evaluation showed that while the effect of the MUR power uprate on the step load transient increased the alternating stress significantly (by a factor of 2.5) at the nozzle, the original UF was calculated too conservatively.

The number of occurrences used in the AOR to calculate the UF for this transient was 34,470 (which is the number of occurrences for heatup/cooldown) instead of the 2,000 occurrences specified for design for step load increases or decreases.

By removing that conservatism and adding the MUR power uprate effect, the UF was reduced from 0.716 (pre-uprate conditions) to 0.333 (the MUR power uprate conditions), and the new CUF was reduced to 0.38 1.* As a result, this component remains acceptable for the MUR power uprate.It is, therefore, concluded that all pressurizer components meet the stress and fatigue analysis requirement of Section III of the ASME Code 1965 Edition, up to and including the winter 1967 Addenda for plant operation at the MUR power uprate conditions.

1V.3.7.2 Pressurizer Surge Line Piping Parameters associated with the MUR power uprate were reviewed for their impact on the design basis analysis for the pressurizer surge line piping including the effects of thermal stratification.

Nuclear Steam Supply System design parameters, NSSS design transients, and changes at the reactor coolant loop auxiliary Class 1 branch nozzle connections due to deadweight, thermal, seismic, and LOCA loading conditions were considered.

Thermal stratification takes place during plant transients (e.g., during plant heatup), and the temperature ranges defined in the stratification AOR were conservatively based on plant operating data. Thot has increased slightly for the MUR power uprate (see Table IV-1). This change has a negligible effect on the 49 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION stratification AOR, since it only results in a slight reduction in the delta-T between the pressurizer and the hot leg during steady-state normal operation.

Therefore, the stratification temperature ranges developed in the AOR bound the new operating conditions.

There is no impact on the deadweight analysis due to the MUR power uprate because there is no discernable change in the weight of the auxiliary Class I pressurizer surge line piping systems. Fluid weight changes due to the change in Thor are very small, and their effect on the overall piping weight is insignificant.

The seismic response spectra remain unchanged.

Therefore, there is no impact on the seismic analysis.

Although BLPBs could be invoked through LBB implementation, continuing to base the RCS structural analyses on the original design basis LOCA events is conservative

'and valid.Therefore, no change to the LOCA hydraulic forcing functions is required.

In conclusion, the MUR power uprate has no impact on auxiliary Class 1 branch nozzle connection loads resulting from the deadweight, thermal, seismic, or LOCA input loading conditions.

It is noted in the introduction to this section, however, that some of the NSSS calculated thermal transients are affected by the MUR power uprate. The calculated transients refer to thermal transients re-calculated for the MUR power uprate conditions.

Reconciliation between the design transients and re-calculated transients for the MUR power uprate was performed for critical locations on the RCS surge lines and fittings.

These reconciliations are summarized below.Surge Line Piping The critical margin on the maximum primary-plus-secondary stress is 1.78%. Per the AOR, the maximum calculated stress intensity range is based on design pressure, dead weight, seismic loads, and specified normal operating transients for surge line piping.The design pressure, dead weight, and seismic loads are not affected by the MUR power uprate. Per the AOR, the dominant stresses are from the plant loading transient and the plant unloading transient.

The effect of the MUR power uprate on these transients was evaluated.

Based on this evaluation, these transients as originally specified for design remain applicable for the MUR power uprate. Therefore, the stresses on the surge nozzle are also unchanged.

The critical CUF of 0.937 in the surge line occurs at the elbow under the pressurizer.

This CUF is primarily due to stratified flow and striping, was developed for the Combustion Engineering Owners'Group (CEOG), and represents a bounding case for the combined effects of stratified flow and striping on the maximum CUF of any CE plant surge line.The report to the CEOG (Reference IV-1) stated the following with respect to the calculated generic CUF of 0.937: The actual usage factor for each specific plant is expected to be lower because 1) the loadings are generic and very conservative, 2) the assumptions made for material properties are conservative, and 3) the most highly stressed line (elastically) was used as the line for shakedown.

The highest contribution to fatigue results from a load set which ranges between a non-stratified load state and a 340 °F stratified flow load state. Virtually all of the cumulative usage includes load sets with a stratified flow load state. This indicates that the OBE and full flow thermal stresses contribute very little to the overall fatigue conclusions.

Therefore, there is a much greater margin on the allowable CUF for the Calvert Cliffs surge lines, and the actual margin for the Calvert Cliffs surge lines is not considered critical.

As documented in the piping specification, all transients affecting the RCS piping, including the surge line, are unchanged by the MUR 50 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION power uprate. Consequently, the Calvert Cliffs surge line piping remains acceptable for the MUR power uprate.Surge Line Temperature Measurement Nozzle The surge line temperature measuring nozzle is a RTD nozzle that has a pre-uprate CUF of 0.732, and the transient contributing the largest UF (0.333) is the LOSP transient in combination with heatup, neither of which is affected by the MUR power uprate conditions.

The design transients contributing to the RTD nozzle fatigue usage that are also affected by the MUR power uprate are reactor trip, loss of flow, step load increase/decrease, and plant unloading.

The step load increase/decrease transients under the MUR power uprate conditions will not increase the alternating stresses or UF on the RTD nozzle. Therefore, the delta UF calculation for the MUR power uprate includes only the effects from the reactor trip, loss of flow, and plant unloading transients.

Using a conservative 35 years of additional operation to end-of-life in order to envelop plant operation under the MUR power uprate conditions, the CUF is increased by 0.109, from 0.732 to 0.841. This CUF continues to meet the acceptance criterion.

The surge line temperature measuring RTD nozzle is therefore considered acceptable for the MUR power uprate.Surge Line Sampling Nozzle The surge line sampling nozzle has a pre-uprate CUF of 0.996, and the transient contributing the largest UF (0.263) is the LOSP transient in combination with heatup, neither of which is affected by the MUR power uprate conditions.

In addition, this UF was very conservatively generated originally using the simplified elastic-plastic analysis (i.e., application of the K, factor as defined in Paragraph NB-3228.5 of the ASME Code). A full elastic-plastic analysis reduces this UF from 0.263 to 0.044, thereby reducing the pre-uprate CUF to 0.777 (0.996 -0.263 + 0.044).The design transients that contribute to the sampling nozzle fatigue usage and that are also affected by the MUR power uprate conditions are reactor trip, loss of flow, step load increase/decrease and plant unloading.

The step load increase/decrease transients under the MUR power uprate conditions will not increase the alternating stresses or UF on the sampling nozzle. Therefore, the delta UF calculation for the MUR power uprate includes only the effects from the reactor trip, loss of flow, and plant unloading transients.

Using a conservative 35 years of additional operation to end-of-life in order to envelop plant operation under the MUR power uprate conditions, the CUF is increased by 0.207, from 0.777 to 0.984.This CUF continues to meet the acceptance criterion.

The surge line sampling nozzle is, therefore, considered acceptable for the MUR power uprate.Based on the above, the existingpressurizer surge line piping analysis remains valid.IV,4 EFFECTS OF OPERATING POINT DATA VARIATIONS The M1UR power uprate operating point values shown in Table TV-i represent a best estimate.

In all probability, the MUR power uprate operating point may move slightly over time, resulting in small changes in the operating point parameters.

Regardless of these small anticipated changes, particularly in the operating temperatures and the resulting delta-T, the structural AOR performed for the Calvert Cliffs Units I and 2 RCS components remains bounding.

The following discussion is based on the fact that the AOR considered Thot and T 0 old design values of 604'F and 548'F, respectively, with a resulting delta-T of 56'F.51 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION IV.4.1 Reactor Coolant System Thermal Movements The maximum thermal movements of various locations on the RCS (e.g., tributary nozzle ends) result from the change in RCS temperature from ambient conditions to operating conditions.

The MUR power uprate thermal movements will be enveloped by the AOR results, since AOR results are based on ambient to operating condition nominal temperature ranges that bound the temperature ranges associated with the MUR power uprate conditions.

This was demonstrated in an analysis comparing RCS thermal movements due to design operating setpoint temperatures to similar results determined at the MUR power uprate nominal setpoint temperatures.

The MUR power uprate condition thermal movements either remained the same or decreased slightly, relative to the movements due to design operating setpoint temperatures.

In general the decreases were on the order of I to 2%. Maximum decreases were 4 to 5%.Furthermore, this conclusion will remain valid if the nominal values of Thot and Tcold vary slightly after the MUR power uprate, because 1) there is sufficient margin between the MUR power uprate nominal Thor value of 595.9°F and the design Th., of 604'F, and 2) the Tcold value is anticipated to remain at the design value of 5487F, which has been the case for previous plant operation.

IV.4.2 Reactor Coolant System Loads Reactor Coolant System component nozzle and primary piping thermal expansion loads are directly affected by delta-T, the temperature difference between Tht and Tcold. Given the same RCS configuration and operating temperatures that are generally the same, lower delta-T values result in lower piping and nozzle loads, which in turn result in proportionally lower loads at intermediate component locations and at the component supports.

This conclusion can be drawn because the general RCS characteristics of stiffness, mass and connectivity will not change for the MUR power uprate, thus resulting in an overall RCS load distribution for the MUR power uprate conditions that are very similar to the load distribution analyzed in the AOR.The delta-T values associated with current and the MUR power uprate conditions are both less than the delta-T value used in the AOR. Therefore, even though delta-T increases slightly when going from the current to the MUR power uprate conditions (by -I°F), the AOR piping, component and component support thermal expansion loads remain bounding, because they are associated with a higher value of delta-T.Per Section IV.3, the majority of the AOR design thermal transients remain bounding for the MUR power uprate. Even those that do not remain bounding were demonstrated to have little effect on the AOR stress calculations (see detailed discussions in Section IV. 10.3).Original design basis RCS seismic analysis results are negligibly affected by the MUR power uprate, because small changes in temperature have virtually no effect on the material properties of the structure, and therefore, on the manner in which the structure responds to a given set of input loads. Furthermore, Section WV.2 concludes that it is valid to base the MUR power uprate LOCA evaluations on the original DBEs. Furthermore, since LBB can be used to mitigate any adverse effects from the MCLB load contributions, basing the MUR power uprate LOCA evaluations on the original DBEs is both valid and conservative.

Finally, since the RCS structure responds to the same design input loadings in essentially the same manner under the MUR power uprate conditions, the original design basis structural analysis results remain valid.52 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION IV.4.3 Reactor Coolant System Stresses and Usage Factors Since the AOR normal operating -conditions, seismic, and LOCA structural analysis results remain bounding for the MUR power uprate, the only changes to the AOR design, emergency, and faulted condition load combinations used to calculate the stresses and fatigue UFs of record are related to the design thermal transients.

As discussed above and throughout Section IV.3, the CUFs determined in the AOR were insensitive to the effects of the transient input changes associated with the MUR power uprate.It is safe to conclude that any further, even smaller, changes resulting from operating point drift will also be acceptable.

It is also noted that the ASME Code stress allowables used in the AOR are unaffected by small changes in operating temperatures, leading to the conclusion that the bounding stresses determined in the AOR will continue to remain below their corresponding ASME Code allowables.

Consequently, the structural integrity of the RCS components is further confirmed for small variations in the MUR power uprate conditions, and the stress margins identified in the AOR calculations remain applicable.

IV.5 REACTOR VESSEL INTEGRITY The factors influencing RV integrity are the initial properties of the materials and the neutron fluence incident on the materials.

The MUR power uprate does not affect the initial material properties, but the neutron fluence can change. The effect of neutron fluence changes on vessel integrity are assessed below using 10 CFR Part 50, Appendices G and H, and 10 CFR 50.61.Pressurized Thermal Shock -The screening criteria in 10.CFR 50.61 are 2706F for forgings, plates, and axial welds and 300'F for circumferential welds. The highest RTPTs value for Calvert Cliffs Unit I at the end of the extended license was determined to be 255'F which is associated with'the RV lower shell course axial weld seams. This is based on a projected fluence of 5.1 lx10 1 9 n/cm 2 , E>1MeV. The highest RTPTS value for Calvert Cliffs Unit 2 at the end of the extended license was determined to be 199°F which is associated with the RV lower shell course plate D8906-1. This is based on a projected fluence of 5.79x10 1 9 n/cm 2 , E>1MeV. In both cases the projected value of RTrs is less than the pressurized thermail shock screening criterion of 270'F such that the planned uprate does not result in exceeding the screening criterion.

Vessel Fluence Evaluation

-The Units 1 and 2 extended end-of-life neutron fluence values were re-evaluated assuming a 1.4% MUR power uprate (which bounds the requested MUR power uprate of 1.38%)i The fluence results are listed in the following table.Unit 1 Fluence Unit 1 Fluence Unit 2 Fluence Unit 2Fluence1.4% Power Current 1.4% Power Uprate Uprate Critical Weld 5.09E+ 19 5.11 E+ 19 5.74E+ 19 5.79E+ 19 1/4 T Location 3.06E+ 19 3.08E+19 3.02E+ 19 3.05E+ 19 3/4 T Location 6.09E+ 18 6.12E÷ 18 6.33E+18 6.38E+ 18 For completeness, the RTPTS values for the RV welds and plates are listed in the following table. Note that the RTPTS values are well below the 10 CFR 50.61 pressurized thermal shock screening criteria limits of 270'F for plates, forgings, and axial weld materials, and 300'F for circumferential weld materials.

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION

__ __ _ ___Unit I Unit I Unit 1 Unit I{.Uniti Uniti 1.4% Power 1.4% Power Current Current... * ....Uprate Uprate Seam/Plate Fluence n/cm 2 RTPTS 'F Fluence n/cm 2 RTPTS 'F WELDS 2-203-A/B/C 5.09E+19 243.6 5.11 E+19 243.7 3-203-A/B/C 5.09E+19 254.1 5.11 E+19 254.2 9-203 5.09E+19 53.4 5.11E+19 53.5 PLATES D-7206-1 5.09E+19 158.0 5.11E+19 158.1 D-7206-2 5.09E+ 19 105.1 5.11 E+19 105.2 D-7206-3 5.09E+19 145.1 5.11E+19 145.2 D-7207-1 5.09E+19 170.5 5.11E+19 170.6 D-7207-2 5.09E+19 139.3 5.1 IE+19 139.3 D-7207-3 5.09E+19 118.0 5.11E+19 118.1 Unit 2 Unit 2 Unit 2 Unit 2 S .:1.4% Power 1.4% Power Current Current:

Uprate Seam/Plate Fluence n/cm 2 RTPTS OF Fluence n/cm 2 RTPTS 'F WELDS 2-203-A/B/C 5.74E+19 122.9 5.79E+19 123.0 3-203-A/B/C 5.74E+19 55.1 5.79E+19 55.3 9-203 5.74E+ 19 72.3 5.79E+ 19 72.4 PLATES D-8906-1 5.74E+ 19 198.3 5.79E+ 19 198.5 D-8906-2 5.74E+19 149.7 5.79E+19 149.8 D-8906-3 5.74E+19 179.0 5.79E+19 179.2 D-8907-1 5.74E+19 183.1 5.79E+ 19 183.3 D-8907-2 5.74E+19 167.0 5.79E+ 19 167.1 D-8907-3 5.74E+19 128.0 5.79E+19 128.1 Heatup and Cooldown Pressure Temperature Limit Curves and Low Temperature Overpressure Protection

-10 CFR Part 50, Appendix G addresses the limits on pressure and temperature that are placed on heatup and cooldown during normal operation.

There are no significant changes to the values used to establish the Appendix G normal operating limits. The 0.05x10'9 n/cm 2 increase in fluence results in less than 0.3°F change to the adjusted reference temperature at the one-quarter thickness location.

The low temperature overpressure protection limits for the MUR power uprate conditions are unchanged for those same reasons.Upper Shelf Energy -10 CFR Part 50, Appendix G requires that the upper shelf energy throughout the life of the vessel be no less than 50 ft-lb. Projections were done in accordance with Regulatory Guide 1.99, Revision 2, and were based on the neutron fluence values through the end of the extended license adjusted to represent conditions for power uprate. For Calvert Cliffs Units I and 2, the upper shelf energy values at the end of the current license were determined to range from 52 ft-lb to 85 ft-lb for the RV 54 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION beitline plates and welds. This demonstrates that all the beitline materials will exceed the upper shelf energy screening criteria.Surveillance Capsule Withdrawal Schedule -10 CFR Part 50, Appendix H defines the RV surveillance program that is to be used by the licensee to monitor the neutron radiation induced changes in fracture toughness of the vessel during the life of the plant. It includes requirements to establish a surveillance capsule withdrawal schedule.

The schedule for Calvert Cliffs Units I and 2 has been updated based on the fluence projected for the extended license. The vessel fluence is predicted to increase only 0.04x10 1 9 n/cm 2 , E>IMeV, as a result of the planned uprate. Therefore, the updated surveillance capsule withdrawal schedule is also applicable under conditions including the MIJR power uprate.IV.6 NUCLEAR FUEL This section summarizes the evaluations performed to determine the effect of the MUR power uprate on the nuclear fuel. The core design for Calvert Cliffs is performed on a fuel cycle specific basis and varies according to the needs and specifications for each fuel cycle. However, some fuel-related analyses are not cycle specific.

The nuclear fuel review for the MUR power uprate evaluated the fuel assembly mechanical performance, the fuel core design, thermal-hydraulic design, and fuel rod performance.

IV.6.1 Fuel Assembly Mechanical Performance The Calvert Cliffs 14x14 fuel design was evaluated to determine the impact of the MUR power uprate on the fuel assembly design criteria.

The evaluation concluded that the Calvert Cliffs fuel design remains acceptable and continues to satisfy the required design criteria under the operating temperature, operating pressure, and flow rates resulting from the MUR power uprate conditions.

The evaluation methodology compared significant operating parameter values, used in the AOR, with the values of those same parameters proposed for the MUR power uprate. The significant parameters evaluated included inlet temperature, system pressure, core average LHRs, maximum fuel rod axial average fluence, minimum coolant flow rate, fuel residence time, and peak fuel rod burnup. These parameters affect such important design criteria issues as the fuel rod stress, strain, fatigue, and clad collapse, as well as the fuel assembly hold-down margin, and shoulder gap. The evaluation of the comparison of these significant parameter values showed that the proposed MUR power uprate operating and transient values are the same as or bounded by the existing AOR values except for the core average LHR. Sufficient margins exist, however, to-allow for the power uprate increase in that parameter.

Since the core plate motions for the seismic and LOCA evaluations are not affected by the uprated conditions, there is no impact on the fuel assembly seismic/LOCA structural evaluation.

Therefore, the fuel mechanical performance design criteria will continue to be satisfied under the proposed MUR power uprate conditions.

IV.6.2 Fuel Core Design The impact of a bounding 1.7% uprate condition on the fuel core design was evaluated against the data used in the current Calvert Cliffs safety AOR. Since the MUR power uprate is relatively small, the range of parameters used in the current safety AOR are adequate to accommodate the range of parameters expected for future cores that have implemented the MUR power uprate. The core analyses for specific uprate cycles has shown that the implementation of the MUR power uprate does not result in significant changes to the current nuclear design basis for the safety analysis documented in the UFSAR. The impact of the MUR power uprate on peaking factors, rod worths, reactivity coefficients, shutdown margin, and kinetics parameters is either well within normal cycle-to-cycle variation of these values or controlled by the core design and will be addressed on a cycle-specific basis consistent with reload methodology.

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION The methods and core models used in the MUR power uprate analyses are consistent with those presented in the Calvert Cliffs UFSAR. No changes to the nuclear design philosophy, methods, or models are necessary due to the MUR power uprate. The current range of required cycle specific analysis is sufficient to verify the applicability of these parameters for future cycles.IV.6.3 Core Thermal-Hydraulic Design The core thermal-hydraulic design and methodology were evaluated for the MUR power uprate. The thermal hydraulic design is based on the TORC computer code described in Reference IV-2, the ABB-TV and ABB-NV [non-mixing vane] critical heat flux correlations described in Reference 1V-3, the simplified TORC modeling methods described in Reference IV-4, and the CETOP-D code described in Reference TV-5. In addition, the DNBR analysis uses the methodology for determining the limiting fuel assembly or assemblies.

The Extended Statistical Combination of Uncertainties presented in Reference IV-6 and approved in Reference IV-7 was applied to validate the design limit of 1.24 on the ABB-TV and ABB-NV minimum DNBR. This DNBR limit includes the following allowances:

1. NRC specified allowances for TORC code uncertainty.

2. Rod bow penalty equivalent to 0.6% on minimum DNBR as discussed in Reference IV-8.The core thermal-hydraulic design and methodology remain applicable for Calvert Cliffs with the MUR power uprate.IV.6.4 Fuel Rod Design As noted in previous sections (e.g., 11.3) Calvert Cliffs Unit I Cycle 17 was originally targeted as the lead unit for implementation of the MUR power uprate. Subsequent to performance of the Unit I Cycle 17 analyses and evaluations, a fuel design change that implements the use of ZrB 2 integral burnable absorbers was submitted to the NRC (Reference W-17) and approved (Reference IV-18). Unit 2 Cycle 16 was the lead unit for the fuel design change. The analyses performed to support the transition to ZrB 2 have already included the MUR power uprate as an initial condition.

No additional analyses to support both ZrB 2 and the MUR power uprate are required.

Application of the MUR power uprate analyses and evaluations have been included as part of the normal reload process for all subsequent Calvert Cliffs Units I and 2 cores since that time (presently Unit 1 is on Cycle 19 and Unit 2 is on Cycle 17).Starting with the Calvert Cliffs Unit 1 Cycle 17 core, the thermal performance of Erbia and U0 2 fuel rods with the MUR power uprate were evaluated using the FATES3B version of the fuel EM (References IV-7, IV-8, and IV-9), the Erbia burnable absorber methodology described in Reference IV-10, the maximum pressure methodology described in Reference IV- 11, and the ZIRLOTM fuel rod cladding methodology described in Reference IV-12. This evaluation included a power history that enveloped the power and burnup levels expected for the peak fuel rod at each burnup interval, from beginning-of-cycle to end-of-cycle burnups, including a reduction in maximum permitted Fr, consistent with implementation of the power uprate. The maximum predicted fuel rod internal pressure for the uprated core remains below the critical pressure for No-Clad-Lift-Off (Reference IV- 14).The thermal performance of Erbia and U0 2 fuel rods for Calvert Cliffs Unit 1 (subsequent to Cycle 17)and Unit 2 (subsequent to Cycle 15) cores with the MUR power uprate is evaluated using the FATES3B version of the fuel EM (References IV-9, IV-10, and IV- 11), the Erbia burnable absorber methodology described in Reference IV-12, the maximum pressure methodology described in Reference IV-13, and the 56 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION ZIRLOTM fuel rod cladding methodology described in Reference IV-14. These evaluations include a power history that envelopes the power and burnup levels expected for the peak fuel rod at each burnup interval, from beginning-of-cycle to end-of-cycle burnups. The maximum predicted fuel rod internal pressure for the uprated conditions will be shown to remain below the critical pressure for No-Clad-Lift-Off (Reference IV- 13).The expected fuel rod corrosion performance for Calvert Cliffs Units 1 and 2 cores with MUR power uprate was evaluated and found acceptable.

This evaluation was conducted consistent with requirements of the NRC SER on the high burnup topical report for 14x14 CE design fuel of Reference IV-15. This evaluation also considered the impact of recent high duty corrosion observations for OPTINTM clad fuel that may be resident in Calvert Cliffs Units 1 and 2 (see also Reference fV-16). The fuel rod corrosion performance of OPTINTM and ZIRLOTM clad fuel specifically for the Calvert Cliffs Unit I Cycle 17 core (including MIUR power uprate) were evaluated (References IV-14, IV-15, and IV-16) and found to be acceptable.

The fuel rod corrosion performance for MIUR power uprated Calvert Cliffs Unit I (subsequent to Cycle 17) and Unit 2 (subsequent to Cycle 15) cores is evaluated using this same methodology.

IV.7 BALANCE OF PLANT PIPING The balance of plant (BOP) piping systems impacted by the uprate (main steam, feedwater, extraction steam, moisture separator drains, reheater drains, condensate, and heater drain piping) have been evaluated by comparing the conditions for the proposed power uprate with the current operating conditions.

The design temperatures and pressures used in the analyses continue to bound the uprate conditions.

The maximum operating temperatures with the MUR power uprate are within 1% of the existing maximum operating temperatures.

The BOP piping systems remain acceptable for operation at the MUR power uprate conditions, and the proposed 1.38% power uprate will not have adverse effects on the BOP piping.IV.8 CODE OF RECORD The allowable stress formulae defining the primary stress limits for the core shroud, as specified in the Calvert Cliffs UFSAR, were adopted prior to the establishment, of specific design criteria for core support structures by the ASME Boiler & Pressure Vessel Code. Core support structure-specific design criteria were formally introduced as Subsection NG in the Winter 1973 Addendum to Section III of the ASME Boiler & Pressure Vessel Code. Therefore, the core shroud evaluation described above, used allowable stress values defined in Subsection NG of the Winter 1973 Addendum.

Rules for the evaluation of core support structures at elevated temperatures have not yet been approved.

Subsection NH in the 1998 Edition of Section III of the ASME Code, which provides rules for the design of Class I components in elevated temperature service, was therefore used to adjust the allowable stress values defined in Subsection NG of the Winter 1973 Addendum.In conclusion, the Calvert Cliffs Units 1 and 2 primary piping and tributary nozzles remain within allowable stress limits in accordance with ASME Section III, 1965 edition, up to and including the Winter 1967 Addendum (and in accordance with ASME Section III, 1986 Edition for components with MNSA).Furthermore, no piping or pipe restraint modifications are required as a result of the increased power level, because conservatively determined LOCA loads due to MCLBs were used to design the pipe restraint systems.57 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION IV.9 COMPONENT INSPECTION, TESTING, AND EROSION/CORROSION PROGRAMS IV.9.1 Flow Accelerated Corrosion The MUR power uprate has no immediate impact on the flow accelerated corrosion program scope but does result in a slight increase in long-term scope. This long-term impact includes increased inspection scope for some specific systems and possibly some additional replacement scope prior to the end-of-plant life expectancy.

The components included in the increased inspection scope will be determined by analyzing the projected wear rate changes through the use of Chec-Works modeling software.

It is expected that the feedwater system would experience the largest increase in wear. However, it should be noted that, even in the feedwater line, the wear rate changes from the MUR may be undetectable using measurement techniques.

This is due to the fact that velocity changes are predicted to be minimal, thereby causing little change in wear rates experienced by the systems.IV.9.2 Inservice Inspection Program The inservice inspection (ISI) program defines the scope and method of examination of Class 1, 2, and 3 components, and also supports the procedures and examination schedule of these components at Calvert Cliffs.The MUR power uprate does not impact the scope, method of examination, schedule and requirements, or criteria of the ISI program. Additionally, the operating condition changes associated with the MUR power uprate are bounded by the design of the ISI components and supports and do not affect the program scope, selection criteria, or acceptance standards.

Therefore, the ISI program is not affected by the MUR power uprate.IV.9.3 Inservice Testing Program The inservice testing (IST) program at Calvert Cliffs defines the scope of Class 1, 2, and 3 pumps and valves to be tested, the test method, and test schedule.The MUR power uprate does not impact the scope, test methods, schedule and requirements, or criteria of the IST program. Additionally, the operating condition changes associated with the MUR power uprate are bounded by the design of the IST pumps and valves and do not affect the scope, selection criteria, or acceptance standards.

Therefore, the IST program is not affected by the MUR power uprate.IV.9.4 Alloy 600 Program Industry experience in PWRs has shown that Alloy 600 (Inconel 600) components and Alloy 82/182 weld filler metals are susceptible to primary water stress corrosion cracking.

Theprogram includes all Alloy 600 components and Alloy 82/182 welds that are part of the RCS pressure boundary, integral attachments to the RCS pressure boundary, or can have a direct or indirect effect on the integrity of the RCS pressure boundary.

These components include: partial penetration welded nozzles and penetrations in the RCS fabricated from Alloy 600 material, welds made with Alloy 82 or 182 filler metal, full-penetration welds made with Alloy 82 and 182 filler metal, and Alloy 600 piping components, non-pressure boundary Alloy 600 components such as welded internal attachments to vessels, and thermal sleeves. Steam generator tubes and the associated tube-to-tube sheet seal welds, are specifically excluded from this program.This program has assessed the Alloy 600 components and for each of them has documented the risk of component failure. As part of the program, system reliability is evaluated with respect to potential for equipment degradation.

The system reliability is in part based on Calvert Cliffs susceptibility modeling of Alloy 600 components.

Primary water stress corrosion cracking has been shown to be predominantly temperature and environment dependent.

As such, with an increase in RCS temperature, Alloy 600 58 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION susceptibility could potentially be challenged.

Therefore, a review was performed on the impact of a temperature increase as a result of the MUR power uprate with regards to Alloy 600 susceptibility.

As part of the MUR power uprate it was determined that the RCS temperature would only increase by 0.87F on the hot leg. The RCS pressure, flow, and cold leg temperatures would remain the same. Thus, it is anticipated for the worst case scenario that the overall increase experienced by Alloy 600 materials is a 0.87F increase.

The review of this increase on the Alloy 600 components concluded that this increase in temperature affects the Alloy 600 component aging but has an insignificant impact on the components' risk of failure.IV.9.5 Coatings Coatings used within the Containment were specified based on their ability to withstand accident conditions.

The Containment is designed to withstand an internal pressure of 50 psig at 2767F including all thermal loads resulting from the temperature associated with this pressure (UFSAR Section 14.20.2).The coatings within the Containment are not impacted by the MUR power uprate since the mass and energy values are not changed from previously analyzed conditions.

IV.9.6 Steam Generator Program The purpose of the SG program is to ensure the structural and leakage integrity of the tubes through the implementation of the following program elements: " Assessment of existing degradation mechanisms in the reactor coolant pressure boundary within the SG* Steam generator inspection in accordance with the Electric Power Research Institute PWR SG examination guidelines" Assessment of tube integrity after each SG inspection to ensure that the performance criteria for the operating period have been met and will continue to be met for the next period* Maintenance, plugging, and repairs of SG tubes* Primary-to-secondary leakage monitoring" Maintenance of SG secondary side integrity" Primary side and secondary side water chemistry* Foreign material exclusion* Self-assessment of the SG program 0 Preparation of NRC and industry reports A review of the SG program elements has concluded that the program elements are symptom based, augmented by regular inspections, maintenance and chemistry activities, and industry experiences.

At the MUR power uprate conditions, the SG tubes are exposed to a 0.87F increase in temperature.

This temperature increase slightly increases the chance of stress corrosion cracking in the SG tubes. The existing plugging margin and inspection program elements are sufficient to ensure tube integrity.

The SG program elements are independent of the reactor thermal power and therefore, the SG program elements are not affected by the MUR power uprate.59 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION IV.9.7 Containment Leak Rate Program The containment leak rate testing program performs the Type A, B, and C containment leakage testing to verify the integrity of the Containment and those systems and components which penetrate the containment walls.The MUR power uprate does not impact the scope, requirements or criteria of the containment leak rate testing program. Additionally, the operating condition changes associated with the MUR power uprate do not affect the Containment or the systems and components which penetrate the containment walls. The containment pressure following a DBA from the MUR power uprate conditions is bounded by the AOR performed at 102% thermal power. Therefore, the containment leak rate testing program is not affected by the MUR power uprate.IV.9.8 Motor-Operated Valve Program The proposed MUR power uprate will not impact the Generic Letter 89-10 Motor-Operated Valve (MOV)program. The following systems contain valves within the MOV program: 1. Instrument air system, 2. Safety injection system, 3. Plant drain system, 4. Primary containment heating and ventilation, 5. Containment spray system, 6. Chemical and Volume Control System, 7. Reactor Coolant System, and 8. Feedwater system.The variables that could affect MOV performance are increased differential pressure across the valve, increased effects of pressure locking/thermal binding, and increased temperature experienced by the actuator motor.Systems I through 5 above are not impacted by the MUR power uprate. The 11 valves in systems 6 through 8 could potentially be impacted by the uprate. The differential pressure calculations for these valves were reviewed and all of them use system design pressures for calculating maximum differential pressure.

Since the system design pressures are not changing there is no effect on the calculated differential pressures across the valves.The valves susceptible to pressure locking or thermal binding are the power-operated relief block valves and the SDC return line valves. These valves possess engineered features that preclude pressure locking or thermal binding.The maximum design temperature of the room in which the motor is located is used to calculate the torque reduction effect of increased temperature.

Since the design temperatures are not changing there will be no effect on the MOV motors.As identified in Section 11.1, there are no changes to the safety analysis (i.e., existing analysis of the MSLB and LOCA remain bounding).

Consequently, this results in no impact to the MOV program.60 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION IV.9.9 Air-Operated Valves The air-operated valve (AOV) program was evaluated for impact due to the proposed MUR power uprate.A review considered valves in the main steam, feedwater, and other secondary side systems. Valves in primary side systems such as SI and CVCS should not be affected since RCS pressure is not changing, From an AOV program standpoint, the main concern is differential pressure across the valve and flow through the valve. Thrust calculations to determine required outputs from the air-operators conservatively assume worst case differential pressure across a valve. The following valves are addressed in the review: I. Atmospheric Dump Valves 2. Feedwater Regulating Valves 3. Feedwater Regulating Bypass Valves 4. Main Steam to Auxiliary Feedwater (AFW) Pumps There is no impact from a thrust standpoint because the calculations assume the highest SG pressure based on pressure limits, MSSV settings, and SG feed pumps running at shutoff head. These are conservatively higher pressures or D/Ps than the MUR power uprate will implement, so there is no impact on actuator capability.

Based on the information that was provided through the heat balance calculation generated using the plant specific model, there is no impact to AOV program valves from an actuator thrust standpoint.

IV.9.10 Non-Program Valves Since increases in differential pressure are minimal and flow rates will only increase about 2% or less, it was determined that the control valves will be able to handle the increased flow due to the MUR power uprate.IV.10 FIRE PROTECTION This evaluation has been conducted in order to evaluate the effects of the MUR power uprate on the plant's fire protection program.The plant fire protection program is the integrated effort involving systems, structures, components, procedures, and personnel used to carry out all activities of fire protection, fire prevention, and to ensure safe shutdown following a fire event. The fire protection program uses a defense in depth concept to prevent fires from starting, to detect, control, and suppress those fires that do occur; and to ensure that fire will not prevent essential plant safety functions from being performed.

Both units are served by a fire protection system that provides a reliable fire protection water supply delivering fire protection water in quantities sufficient to satisfy the maximum probable demand; and, automatic and manual fire protection systems and equipment that provide fire suppression capabilities.

The fire protection program and fire protection features are described in Calvert Cliffs UFSAR Section 9.9. Fire protection features include the fire water supply; fire pumps and distribution piping;fixed water suppression systems; fixed gaseous suppression systems; manual fire suppression systems;and fire detection and alarm systems. Passive fire protection features include fire barriers and fire rated penetration seals. Fire and emergency response activities are performed by the on site fire brigade.61 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION The changes that will occur as a result of the MUR power uprate which increase the thermal and electrical power of the plant have been evaluated with respect to their impact on plant fire protection.

The results of the evaluation are that the MUR power uprate has no affect on the plant's fire protection program.IV.11 APPENDIX R The goal of 10 CFR Part 50, Appendix R, is to ensure safe shutdown of the reactor following a fire in any plant area, thereby preventing core damage and protecting the public. Appendix R applies to plants licensed prior to January 1, 1979.Appendix R compliance can be affected by adding heat to plant areas that could affect Appendix R safe shutdown because the higher temperatures could affect Appendix R equipment and plant operators.

However, the overall temperature changes in the primary and secondary systems are very small such that the issue of added heat load to the plant is not a concern.Appendix R can be affected by .additional decay heat due to the higher power levels. This additional decay heat associated with the changes from the MUR power uprate was evaluated and found to be negligible.

IV.12 HIGH ENERGY LINE BREAK The Calvert Cliffs high energy line break (HELB) analysis was reviewed in support of the MUR power uprate. The activities, elements, and philosophy that currently constitute the HELB analysis are not affected by the MUR power uprate. The slight lowering of the secondary pressure limits the mass flowrate through the break location.

Although an extremely slight increase in enthalpy occurs with a decrease in saturated steam pressure and temperature, the lowered choked flowrate more than compensates for this. As a result, the overall impact from the proposed MUR power uprate is bounded by the existing HELB analysis.

In accordance with Calvert Cliffs design change process, the design change package for installing the Caldon LEFM CheckPlus system was evaluated against the HELB analysis requirements.

No new piping was added, no postulated break locations changed, and no changes were made to the assumed blowdown from any currently-postulated breaks; therefore, there is no impact on the current Calvert Cliffs HELB analysis.IV.13 REFERENCES IV-1 CEN-387-P, Revision 1-P-A, "Pressurizer Surge Line Flow Stratification Evaluation," May 1994 IV-2 CENPD-206-P-A, "TORC Code: Verification and Simplified Modeling Methods," June 1981 IV-3 CENPD-387-P-A, "ABB Critical Heat Flux Correlations for PWR Fuel," May 2000 IV-4 CENPD-206-P-A, "TORC Code: Verification and Simplified Modeling Methods," June 1981 IV-5 CEN-191(B)-P, "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units I and 2," December 1981 IV-6 CEN-348(B)-P-A, Supplement 1-P-A, "Extended Statistical Combination of Uncertainties," January 1997 1V-7 Letter from G.M. Holahan (NRC) to S.A. Toelle (ABB), dated August 31, 1994, "Generic Approval of CEN-348(B)-P-A, 'Extended Statistical Combination of Uncertainties' (TAC No. M90019)" P1-8 CENPD-225-P-A, "Fuel and Poison Rod Bowing," June 1983 WV-9 CEN-161(1B)-P Supplement 1-P-A, "Improvements to Fuel Evaluation Model," January 1992 62 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION IV-10 CENPD-139-P-A, "Fuel Evaluation Model," July 1974 IV- 1 CEN-16 1(B)-P-A, "Improvements to Fuel Evaluation Model," August 1989 IV-12 CENPD-382-P-A, "Methodology for Core Designs Containing Erbium Burnable Absorbers," August 1993 IV-13 CEN-372-P-A, "Fuel Rod Maximum Allowable Gas Pressure," May 1990 IV-14 CENPD-404-P-A, Revision 0, "Implementation of ZIRLOTM Cladding Material in CE Nuclear Power Fuel Assembly Designs," November 2001 IV-15 CEN-382(B)-P-A, "Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/kgU for Combustion Engineering 14x 14 PWR Fuel," August 1993 WV-16 CENPD-384-P, "Report on the Continued Applicability of 60 MWD/kgU for ABB Combustion Engineering PWR Fuel," September 1995 1V-17 Letter from B.S. Montgomery (CCNPP) to Document Control Desk (NRC), dated July 15, 2004,"License Amendment Request: Incorporate Methodology References for the Implementation of PHOENIX-P, ANC, PARAGON, and Zirconium Diboride into the Technical Specifications" IV-18 Letter from R.V. Guzman (NRC) to G. Vanderheyden (CCNPP), dated February 24, 2005, "Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 -Amendment RE: Incorporating Core Operating Limits Analytical Methodology' References into Technical Specifications (TAC Nos. MC4019 and MC4020)" 63 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION V. ELECTRICAL EQUIPMENT DESIGN This section summarizes the evaluations performed to determine the effect of the MUR power uprate on the electrical equipment.

The electrical equipment included in the evaluation is presented within each subsection.

V.1 EMERGENCY DIESEL GENERATORS/STATION BLACKOUT EQUIPMENT Calvert Cliffs onsite electrical distribution systems include non-Class IE plant service transformers and associated busses. The 4.16 kV, 480 Volt, and 120/208 Volt systems include both Class 1E and non-Class lE equipment.

The onsite direct current (DC) distribution system includes both Class 1E and non-Class IE equipment.

The Class IE AC distribution system includes two Class lE 4.16 kV busses per operating unit, each capable of being powered by an associated Class I E standby emergency diesel generator in the event of a loss of offsite power. A station blackout diesel generator is designed to provide sufficient power to any of the four Class IE 4.16 kV busses in order to safely shutdown one unit and maintain it in a safe shutdown condition during a station blackout event. Downstream 480 Volt and 120 Volt busses also feed two trains of redundant safety equipment.

As referenced in Section 11.1, there is no change to the existing accident analyses and they continue to be valid for the MUR power uprate. The electrical motors and supporting equipment are sized for maximum accident load requirements.

Thus, the emergency diesel generators remain sufficient to provide all required electrical loads and there is no need to upgrade any other existing Class I E electrical equipment.

The non-Class lE AC distribution systems provide power for non-safety-related systems during normal plant conditions.

The large non-safety loads powered from these busses include condensate pumps, condensate booster pumps, and heater drain pumps. The MUR power uprate does not result in an increase in mechanical load beyond the design rating of any non-Class 1E equipment.

The motors and associated support and protective equipment are sized based on design ratings, thus they are adequately sized for the small load increase resulting from the MUR power uprate.The onsite DC distribution system will see minor load variations due to the power uprate; however the resulting electrical loads remain within the ratings of the existing distribution system and no changes are required.V.2 MAIN GENERATOR AND ASSOCIATED SYSTEMS The Unit I Main Generator has a design rating of 1,020 MVA at 25 kV 60 Hz when operated at 0.9 lagging power factor (918 MW) and hydrogen pressure of 60 psig. Unit 2 Main Generator is rated at 1,012 MVA at 22 kV 60 Hz when operated at 0.9 lagging power factor (910.8 MW) and hydrogen pressure of 75 psig. The new operating point corresponding to the MUR power uprate is within the design rating of both machines.

The generators are operated to produce power output within the limits of their associated reactive capability curves. If required, the MVAR output of the generator can be adjusted such that the total MVA rating is not exceeded.

No modification to auxiliary or support equipment is required.Applicable calculations were reviewed and determined no changes are required for generator voltage regulator and associated protective relay settings.Two unit transformers are connected via an isolated phase bus to the output of each main generator and are. designed to carry the maximum generator output and transform generator output voltage to transmission system voltage. Each of the paralleled transformers is rated for 810 MVA at 65'C rise. The 64 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION maximum MVA rating of the Units 1 and 2 generators remain at 1020 and 1012 MVA which is well within the rating of the paralleled transformers.

The associated isolated phase bus and switchyard equipment are also rated for maximum current flow from the generator, thus no modification to this equipment is required.

However, the existing unit limitations for conditions when the unit's isophase bus duct cooling is not available or operating at rated capability, or when one of the unit's generator step-up transformers is not available still remain.V.3 ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT In accordance with 10 CFR 50.49, certain electrical equipment must be qualified to operate when exposed to the postulated harsh accident environments of DBAs (i.e., LOCA, MSLB, HELBs). The qualification includes aging considerations of normal plant operating ambient environments.

The effects of the proposed MUR power uprate on the 10 CFR 50.49 EQ program is as follows: V.3.1 Environmental Qualification Accident (Temperature/Pressure)

Environments As discussed in Section 11.5, the current UFSAR Chapter 14 Containment LOCA and MSLB temperature/pressure analyses will not be affected (i.e., remain bounding) considering the MUR power uprate. The current EQ accident (temperature/pressure) environments utilize these UFSAR Chapter 14 analyses.

Therefore, the current inside-Containment EQ equipment LOCA/MSLB temperature/pressure qualification is unaffected by the MUR power uprate.As discussed in Section VII.5, the current UFSAR Chapter 10A outside-Containment HELB temperature/pressure analyses will not be affected (i.e., remain bounding) considering the MUR power uprate. The current EQ accident (temperature/pressure) environments utilize these UFSAR Chapter 10A analyses.

Therefore, the current outside Containment EQ Equipment HELB temperature/pressure qualification is unaffected by the MUR power uprate.V.3.2 Environmental Qualification Accident (Radiation)

Environments As discussed in Section III, the current accident radiation doses, utilized in the EQ program, required re-evaluation as a result of the proposed MUR power uprate. Environmental qualification equipment was re-evaluated against these revised accident radiation doses and confirmed to remain environmentally qualified to these revised accident doses.V.3.3 Environmental Qualification Normal Plant Operating Ambient (Temperature/Humidity)

Environments As discussed in Section VI.6, the heating, ventilation, and air conditioning (HVAC) systems, control normal plant operating ambient environments in Containment and Auxiliary Building.

Environmental qualification equipment is located in both the Containment and the Auxiliary Building.

These HVAC systems were reviewed considering the MUR power uprate. The review determined that these existing HVAC systems will continue to maintain the Containment and Auxiliary Buildings within their UFSAR Chapter 9 design ranges. Therefore, the current normal plant operating aging considerations, utilized in the EQ program, are not impacted and are bounded by the current design.V.3.4 Environmental Qualification Normal Plant Operating Ambient (Radiation)

Environments As discussed in Section III, the current normal operating radiation doses, utilized in the EQ Program, required re-evaluation as a result of the proposed MUR power uprate. Equipment was re-evaluated against these revised normal operating radiation doses and confirmed to remain environmentally qualified to these revised normal operating doses.65 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION V.4 GRID STABILITY The Pennsylvania, New Jersey, Maryland Interconnection has preliminarily reviewed the power uprate for impact on grid stability.

The proposed increase in plant electrical output does not affect the stability of the grid. No switchyard modifications are required as a result of the MUR power uprate.66 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION VI. SYSTEM DESIGN This section presents the results of the evaluations and analyses performed in the NSSS area to support the revised conditions provided previously in Table IV-1. The systems addressed in this section include fluid systems and control systems. The results and conclusions of each evaluation and analysis are presented within each subsection.

VI.A NSSS INTERFACE SYSTEMS VI.I.1 Safety Injection System The function of the SI system is to remove the stored energy and fission product decay heat from the reactor core following a LOCA. The system is designed such that fuel rod damage is minimized, facilitating the long-term removal of decay heat. The system also provides injection.of negative reactivity (boron) in the RCS cooldown events such as a MSLB.The active part of the SI system consists of high pressure SI pumps, the refueling water tank, low pressure safety injection (LPSI) pumps, and the associated valves, instrumentation, and piping.The passive portion of the SI system is the safety injection tanks (SITs) that are connected to each of the RCS cold leg pipes. Each SIT contains borated water under nitrogen pressure and automatically injects into the RCS when the RCS pressure drops below the operating pressure of the SITs. The active portion of the SI system (injection pumps) injects borated water from the refueling water tank into the reactor following a break in either the RCS or steam system piping to cool the core and prevent an uncontrolled return to criticality.

Safety Injection system operation is described in two phases; the injection phase and the recirculation phase. The injection phase provides emergency core cooling and additional negative reactivity immediately following a spectrum of accidents including a LOCA by prompt delivery of borated water to the RV. The recirculation phase provides long-term post-accident cooling by recirculating water from the containment sump.During normal operation the SI system does not operate and has no design function.

Thus, during normal operation, there is no impact on the system due to the MUR power uprate. However, the slight increase in RCS stored energy and decay heat resulting from the power uprate are well within the capabilities of the SI system to respond to DBAs. The results of the evaluation of a LOCA are presented in Section 11.3.For non-LOCA RCS depressurization events, the SI system is acceptable for the proposed power uprate as demonstrated in Section 11.2.VI.1.2 Chemical and Volume Control System The CVCS provides for boric acid addition and removal, chemical additions for corrosion control, reactor coolant cleanup and degasification, reactor coolant makeup, and processing of reactor coolant letdown.During plant operation, reactor coolant letdown is taken from the cold leg on the suction side of the RCP, and is reduced in pressure and temperature prior to it entering the volume control tank. The charging pumps take suction from the volume control tank and return the coolant through the regenerative heat exchanger to the RCS in the cold leg, downstream of the RCP.The nominal TCold for the power uprate remains unchanged at 548'F. As a result, the temperature of the letdown flow is not changed. Consequently, there is no impact on the thermal performance of the CVCS.67 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION The CVCS provides a source of borated water for post-accident injection.

Evaluation of required ECCS water volumes and boric acid concentrations will be performed as part of the normal reload safety evaluation process. The slight increase of N-16 activity at the MUR power uprate conditions has a negligible effect on letdown line decay time requirements.

There will be no change to the letdown and makeup requirements as a result of the MUR power uprate.As previously noted, T 0 old and the reactor coolant mass flow rate remain unchanged.

Increased power is due to a slight increase in Thor and associated increase in Tawe. The increase in Tawe causes a small increase in the makeup requirements for coolant shrinkage during cooldown.

However, this effect is considered negligible, so the system is capable of supporting the MUR power uprate.VI.1.3 Shutdown Cooling System The SDC system is designed to remove sensible and decay heat from the core and to reduce the temperature of the RCS during the second phase of plant cooldown.The SDC system consists of two electrically aligned trains. Each train consists of one heat exchanger, one LPSI pump, associated valves, and instrumentation.

Both trains take suction from a common suction line off one reactor coolant hot leg, and then flow is divided through the LPSI pumps, the tube side of the SDC heat exchangers, and back to the RCS cold legs through a four leg header.The proposed power uprate will affect the SDC system due to an increased heat load from higher decay heat input. Since decay heat is proportional to plant operating power, any increase in RTP will result in an increase in decay heat load.The SDC system was previously evaluated to be capable of supporting the decay heat that would be present based on 102% of 2700 MWt, which is 2754 MWt including uncertainty.

The analytical power level including revised uncertainty with the MUR power uprate remains 2754 MWt. Therefore the system is capable of supporting the MUR power uprate.VI.1.4 Auxiliary Feedwater System The purpose of the AFW system is to provide sufficient feedwater to the SGs for the removal of sensible and decay heat, and to cool the primary system to 300'F in case the condensate or the main feedwater systems are inoperable.

An evaluation was performed to determine whether the current design of the AFW system will satisfy its safety functions and support an MUR power uprate.The AFW and condensate storage tank analyses are based on 102% of 2700 MWt (2754 MWt). The analytical power level, including revised uncertainty, with the MuR power uprate remains 2754 MWt.Therefore, the evaluation concluded that the AFW system and condensate storage tank system are capable of supporting the MUR power uprate.VI.1.5 Main Steam System The MSS is designed to transfer steam from the SGs to the turbine throttle stop valves, the reheaters, and the turbine-driven pumps. The MSS also controls SG pressure by means of steam bypass, dump, or safety valves (high pressure) and MSIVs (low pressure).

The system is designed to accommodate electrical load changes from 15 to 100% power at a rate of 5%per minute and at greater rates over smaller load change increments, up to a step change of 10%. This is normally accomplished by manual CEA movement and adjustment of RCS soluble boron concentration.

The primary impact of the MUR power uprate on the MSS is an increase in main steam flow of about 2%.68 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION There is no change in the MSS operating pressure and temperature.

Steam flow to the high pressure turbine will increase by 2.07% for Unit I and 2.23% for Unit 2. The MSS and associated components were evaluated for the increased steam flow and are capable of supporting the MUR power uprate.However, for Unit 2 there is an economic issue concerning the turbine throttle valves. The Unit 2 turbine throttle valves are currently operating at the valve-wide-open position, which is a limiting factor for taking full advantage of the full MUR power uprate. This limitation does not effect the safe operation of the plant and the necessary hardware changes to eliminate this limitation are addressed as an economic concern.VI.1.5.1 MSSVs Overpressure protection for the shell side of the SGs and the main steam line piping up to the inlet of the turbine stop valve is provided by 16 spring-loaded ASME Code MSSV which discharge to the atmosphere.

Eight of these safety valves are mounted on each of the main steam lines upstream of the steam line isolation valves, but outside the Containment.

The MSSVs are designed for full flow relief pressure of 1085 psig, thereby ensuring that the secondary system pressure is limited to within 110% of its design pressure of 1015 psia during the most severe anticipated system operational transient.

The opening pressure of the valves is set in accordance with ASME Code allowances, with the minimum set pressure at 935 psig, and the maximum set pressure at 1050 psig. The total relieving capacity for all valves on both of the steam lines in either unit is 12.26x 106 lbs/hr of saturated steam (6.13x10 6 lbs/hr per SG). This relief capacity is larger than the steam flow at the MUR power uprate conditions.

The accident analysis shows there is adequate MSSV capability at 102% power.Startup and/or power operation is allowable with MSSVs inoperable within the limitations of the Technical Specifications.

The number of inoperable MSSVs determines the necessary level of reduction in secondary system steam flow and thermal power required by the reduced reactor trip settings of the Power Level-High channels.

The current Technical Specifications were confirmed to remain applicable for supporting the proposed MUR power uprate.VI.1.5.2 MS1Vs One MSIV assembly is provided on each main steam line header in order to protect the reactor and SG from damage due to a rupture in the main steam header down stream of the valves.Closure of the MS1V, within a maximum of six seconds after a trip signal is initiated, prevents rapid flashing and blowdown of water stored in the shell side of the SG, thus avoiding a rapid uncontrolled cooldown of the RCS. Also, the isolation valves prevent release of the contents of the secondary side of both SGs to the Containment in the event of the rupture of one main steam line inside the containment structure.

The MSIVs are not impacted by the MUR power uprate because SG and the MSS operational pressure are not increased., Therefore, the ability of the MSIV to close within the Technical Specification limited closure time following a postulated SLB event is not affected.

The MSIVs are therefore capable to supporting the proposed MUR power uprate.VI.1.5.3 Steam Dump and Bypass System The steam dump and bypass system is used to rapidly remove RCS stored energy and to limit secondary steam pressure following a turbine-reactor trip. The atmospheric steam dump system consists of two automatically actuated ADVs which exhaust to the atmosphere.

The turbine bypass system consists of four turbine bypass valves which exhaust to the main condenser.

The power-operated steam dump valves 69 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION and steam bypass valves obviate opening of the MSSVs following turbine and reactor trips from full power.The system also provides a means of heat removal during hot standby, startup, and during plant cooldown.

The atmospheric steam dump valves are capable of removing reactor decay heat when the condenser is not available.

The total respective capacities of the atmospheric steam dump and turbine bypass valves are nominally 5% and 40% of steam flow with the reactor at full power. This flow is sufficient to control the secondary steam pressure on a turbine trip at the MUR power uprate without necessitating operation of the spring-loaded safety valves. Therefore, the steam dump and bypass system is capable of supporting the proposed MUR power uprate.VI.1.5.4 Main Turbine-Generator The turbine-generator is designed to receive steam from the SGs and convert it into electric energy. The condenser transfers unusable heat to the condenser cooling water and deaerates the condensate.

The closed regenerative turbine cycle heats the condensate and returns it to the SGs.Saturated steam is supplied to the turbine throttle from the SGs through four stop valves and four governor control valves. The steam flows through a dual-flow, high-pressure turbine and then through combination moisture separator-reheaters (two in parallel for Unit 1, four in parallel for Unit 2) to three double-flow, low-pressure turbines which exhaust to the main condenser system.Unit I is a General Electric turbine and Unit 2 is a Westinghouse turbine. The two units are similar in construction and type, and have similar performance characteristics and generating capacity.Each generator has the capability to accept the gross rated output of the turbine at rated steam conditions.

The generator shafts are oil-sealed to prevent hydrogen leakage. Each generator has its own shaft-driven excitation equipment.

The main turbine-generator and their associated components were evaluated for the MUR power uprate conditions.

The primary impact on the main turbines is the increase in main steam flow. Steam flow to the high pressure turbines is expected to increase by 2.07% for Unit 1 and 2.23% for Unit 2. This increase in steam flow is within the design capabilities of each main turbine. The impact on the main generators was previously evaluated in Section V.2. No changes to the main turbine-generators are necessary as they are capable of supporting the MUR power uprate.VI.1.6 Steam Generator Blowdown System Each SG has an upper and lower blowdown line which can be used to control the build-up of soluble and particulate concentrations within the SG. The blowdown system will continue to operate normally with no change at a continuous rate of up to 180 gpm per SG. No changes to SG blowdown are required as the SG blowdown system is capable of supporting the proposed MUR power uprate.VI.1.7 Feedwater and Condensate Systems The feedwater and condensate systems are designed to provide a means for transferring the condensate from the condenser hotwell to the SGs (while at the same time raising the temperature and pressure) and providing a means for controlling the quantity of feedwater into the SGs.70 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION The MUR power uprate results in approximately a 2% increase in condensate flow, due to an increase in extraction steam flow through the feedwater heaters and an increase in the condensate temperature.

A hydraulic calculation was performed verifying the condensate systems ability to perform its design function of delivering condensate from the condenser hotwells to the feedwater system at the required flow, pressure, temperature, and quality at the MJUR power uprate conditions.

The feedwater flow will increase approximately 2% under the MUR power uprate conditions.

A hydraulic calculation was performed to evaluate the feedwater systems ability to provide sufficient flow to the SGs at the M1UR power uprate conditions.

Results of the calculation found the feedwater system capable of providing sufficient flow to the SGs.However, one level of the feedwater heaters on Unit 1 and two levels of feedwater heaters on Unit 2 have been identified as having possible limitations for full M\UR power uprate conditions.

The feedwater heaters are not a safety critical component and may be further evaluated with the potential for replacement/modification as an economic concern.The feedwater and condensate systems are capable of supporting the proposed MUR power uprate.VI.1.8 Extraction Steam System/Heater Drains System The extraction steam and heater drain systems provide a means of heating condensate and feedwater, and for returning condensed steam to the condensate system.The MUR power uprate results in approximately a 2% increase in heater drain flow and a corresponding increase in heater drain temperature, due to the increase in heat load in the feedwater heaters. The system evaluation demonstrated that the equipment can operate at the MUR power uprate conditions, with further action needed to upgrade or evaluate the capability of the Unit 2 heater drain pumps and Unit 2 heater drain tank high level dump valves. These actions are treated as economic issues as they are not safety significant.

The M1UR power uprate results in an increase in extraction steam flows and pressures.

The temperature and pressure ratings for the Units 1 and 2 bleeder trip valves bound the MUR power uprate service conditions based on the maximum working pressures contained in American National Standards Institute (ANSI) B 16.34. The design temperature and pressure ratings for Units I and 2 extraction steam drain trip AOVs and MOVs bound the MUR power uprate service conditions based on the maximum working pressure contained in ANSI B 16.34 and the pressure rating listed on the MOV drawings.The extraction steam and heater drain systems are capable of supporting the proposed MUR power uprate.VI.1.9 Circulating Water System The condensers on both units have an operational limit of a 12'F temperature rise in the circulating water across the condensers.

The MUR power uprate is expected to result in a small increase in the temperature rise across the condenser.

Currently, the actual measured temperature rise is approximately 11.6°F which is expected to rise to approximately 11.8°F after the MUR power uprate. Condenser vacuum in-balance is not adversely affected by the MUR power uprate. The circulating water system is capable of supporting the MUR power uprate.71 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION VI.1.10 Condenser Each unit has one three-shell, single-pass, deaerating-type condenser with divided water boxes. The condenser is capable of condensing the exhaust steam from the main turbine and the SG feed pump turbines under the MUR power uprate plant load. Two of the three condenser shells are connected to discharge lines from the steam dump and bypass system. The condenser is internally equipped to receive the full flow from this system. The condenser is adequately sized and is capable of supporting the proposed MUR power uprate. The condenser will operate with approximately 1.9% higher backpressure under the MUR power uprate during summer conditions

(< 3.8 in-HgA). This is still well below the backpressure limit (5.5 in-HgA).VI.I.11 Heat Balance A plant specific model was developed for each unit for both summer and winter bay temperature limits at both the current and the MUR power uprate conditions.

This detailed model was benchmarked to existing plant operating conditions and used to simulate the estimated impact of the proposed uprate. Output from these simulations (pressure, flows, temperatures) were used as an aid in evaluating the impact the MUR power uprate will have on the plant equipment.

VI.2 CONTAINMENT SYSTEMS VI.2.1 Reactor Coolant System The purpose of the RCS is to remove heat from the core and transfer it to the secondary side of the SGs.The RCS consists of the reactor pressure vessel, two hot leg pipes, two SGs, four RCPs, four cold leg pipes, and one pressurizer with attendant interfacing piping, valves and instrumentation.

Evaluations were performed to ensure that the RCS design basis functions could still be met at the revised operating conditions.

The principal effects of the MUR power-uprate on the RCS are a slight increase in Thot and the increase in decay heat. The normal operating pressure of 2250 psia remains unchanged.

The results of the evaluation of uprated conditions on the RCS functions are described below.a. The increase in Thor will increase the total amount of heat transferred to the MSS. Verification that the major components of the NSSS can support this increase in the normal heat removal function is addressed in this section.b, The increased thermal power can change the transient response of the RCS to normal and postulated DBEs. The acceptability of the RCS with respect to protection functions is addressed in Section II. The acceptability of the RCS with respect .to fatigue evaluations is addressed in Section IV. The setpoints for various control systems will be evaluated for recommended changes prior to plant startup.c. The cold leg temperature remains unchanged at 548'F. As a result, the RCS mass flow is not affected by the MUR power uprate.d, Reactor coolant system design temperature and pressure of 650'F and 2500 psia continue to remain applicable for the uprate conditions.

e, The pressurizer design temperature and pressure of 700'F and 2500 psia continue to remain applicable for the uprate conditions.

f. The pressurizer relief requirements increased slightly due to an increase in RCS stored energy and decay heat. However, the change is well within the relieving capacity of the pressurizer safety valves for the design transient condition (Section 11.2).72 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION The RCS is capable to supporting the proposed MUR power uprate.VI.3 SAFETY-RELATED COOLING WATER SYSTEMS VI.3.1 Service Water System The Service Water (SRW) system is designed to remove heat from the plant's various auxiliary systems.The Saltwater (SW) system provides the cooling medium for the SRW heat exchangers.

System components are rated for maximum duty requirements during normal operation and SDC operation, and are also capable of providing heat removal during a LOCA. The SRW system serves as an intermediate barrier between the various auxiliary systems and the SW system.The turbine plant components cooled by SRW include: a. Generator isolated 3 phase bus duct coolers b. Exciter air coolers c. Generator hydrogen coolers d. Stator liquid coolers (Unit I only)e. Circulating Water System priming pump seal water coolers f. Condenser vacuum pump seal water coolers g. Feed pump turbine lube oil coolers h. Condensate booster pump lube oil and seal water coolers i. Instrument and plant air compressors and aftercoolers

j. Turbine lube oil cooler k. Electro-hydraulic oil coolers 1. Turbine Building sample cooling system m. Seal oil system coolers (Unit 2 only)n. Auxiliary feedwater pump room air cooler The SRW system does not see significant impact with the MUR power uprate. The increased decay heat and turbine auxiliary cooling loads will cause a small increase in the cooling water temperature.

The heat loads increase slightly for the Spent Fuel Pool Cooling (SFPC) in the Auxiliary Building; however, this increase is due to the SDC function, not the MUR power uprate. The impact on the SRW system with the MUR power uprate on the component heat loads has been reviewed.

Some system flow adjustments may be necessary to ensure proper cooling to the affected heat loads, the SRW system has adequate margin to perform its design functions within its design parameters.

As such, the SRW system is capable to supporting the proposed MUR power uprate.VI.3.2 Saltwater System The SW system has three pumps for each unit. The pumps provide the driving head to move SW from the intake structure, through the system, and back to the circulating water discharge conduits.

The system is designed such that each pump has sufficient head and capacity to provide cooling water for the SRW and Component Cooling Water (CCW) systems. The system also cools the ECCS pump room air coolers.The SW system consists of two subsystems in each unit. Each subsystem provides SW to two SRW heat exchangers, a CCW heat exchanger, and the ECCS pump room air cooler in order to transfer heat from those systems to the Chesapeake Bay. Seal water for the circulating water pumps is supplied by both subsystems.

A self-cleaning strainer is installed upstream of each SRW heat exchanger.

Operation of the SW system following a LOCA has two phases: before the recirculation actuation signal and after the recirculation actuation signal. Since the LOCA analysis has been performed at 102% of 73 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION 2700 MWt (2754 MWt) it remains applicable at the MUR power uprate. Therefore, the cooling requirements for both phases are unchanged.

The MUR power uprate results in small increases to the heat loads for the CCW and SRW heat exchangers to be transferred to the SW system, corresponding to a slight increase of the SW discharge piping temperature.

These impacts are negligible on the SW system and component operation.

The margins in the system remain essentially the same as for current operating power levels.VI.3.3 Component Cooling Water System The CCW system is designed to remove heat from the plant's various auxiliary systems. The SW system provides the cooling medium for the CCW heat exchangers.

System components are rated for maximum duty requirements during normal and SDC, and are also capable of providing heat removal during a LOCA. The CCW system serves as an intermediate barrier between the various auxiliary systems and the SW system.The CCW heat exchangers are designed for a CCW supply temperature of 95'F, with a SW cooling supply temperature of 90'F, at normal operating conditions.

Component cooling water may reach as high as 120'F during a LOCA, and during plant cooldown and cold shutdowns.

The MUR power uprate results in a change to the CCW system heat loads. The change has a negligible impact on the CCW system. The increased decay heat has a small impact on the cooling water temperature increase.

Calvert Cliffs has evaluated the most limiting mode of CCW operation at the analytical power level of 2754 MWt, therefore, the CCW system is capable of supporting the proposed MUR power uprate.VI.4 SPENT FUEL POOL COOLING SYSTEM The SFPC system is common to both units. The pool contains water with the proper dissolved concentration of boron and has the capacity to store 1830 fuel assemblies.

The SFPC system is designed to remove the maximum decay heat expected from 1613 fuel assemblies, not including a full core off-load.

The maximum pool temperature in this case is 120'F. The system is also capable of being used in conjunction with the SDC system to remove the maximum expected decay heat load from 1830 fuel assemblies, including a full core discharge.

The maximum spent fuel pool temperature in this case is 130'F.The decay heat source-term used in the evaluation of the SFPC system was determined to be conservative for the proposed MUR power uprate conditions.

Therefore, the SFPC system is capable of supporting the proposed MUR power uprate.VI.5 RADIOACTIVE WASTE SYSTEMS The waste processing systems are designed to provide controlled handling and disposal of radioactive liquid, gaseous, and solid wastes from both units. Design criteria were established to maintain the release of radioactive material from the plant to the environment at levels which are ALARA.The design of the waste processing systems was based upon processing reactor coolant and miscellaneous waste during operation with 1% failed fuel. The proposed MUR power update is for a small power increase and since the radioactive waste processing system is designed to handle 1% failed fuel, the MUR 74 ATTACHMENT (2)

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OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION power uprate does not represent a significant challenge to the liquid or gaseous radwaste processing system.All releases meet the Offsite Dose Calculation Manual (ODCM) limits. By meeting the ODCM limits, the guidelines of 10 CFR Part 50, Appendix I are met. This is confirmed by the effluent data and doses reported to the NRC in the Radioactive Effluent Release Reports required by the Technical Specifications and 10 CFR 50.36a.Therefore, the proposed MUR power uprate has no impact on the radioactive waste system releases.VI.6 ENGINEERED SAFETY FEATURES HEATING, VENTILATION, AND AIR CONDITIONING SYSTEMS The plant ventilating systems are designed to provide a suitable environment for equipment and personnel with a maximum amount of safety and operating convenience.

Potentially contaminated areas are separated from clean areas. Airflow patterns originate in areas of potentially low contamination and progress toward areas of higher activity.

Generally, negative pressures are maintained in potentially contaminated areas and positive pressures in clean areas. The ventilating systems in the Containment, waste processing, and fuel-handling areas are designed for containment of radioactive particles.

The path of the discharge from potentially contaminated areas is directed into the respective plant vent where the radioactivity level is monitored.

The equipment in most critical systems is redundant.

The heat load from the primary systems increases only marginally as a result of the minor change in Thrt.The heat load from the feedwater piping in the Containment, Auxiliary Building (steam tunnel), and Turbine Building were evaluated to account for a < 2°F increase in feedwater process fluid temperature to ensure UFSAR Chapter 9 design basis are not impacted.

The remaining BOP piping temperatures do not change appreciably.

VI.6.1 Containment The Containment is cooled by the containment air coolers. During the summer the air temperature is expected to remain below the 120'F design limit. The total heat load in the Containment during normal operation is calculated to be -7.44 x 106 Btu/hr. The increase of <2.0°F in feedwater temperature could potentially increase the heat load on the cooling system by -400 Btu/hr, clearly inconsequential given the order of magnitude difference considering the original heat load in the building.

This assessment is applicable and valid for both units.VI.6.2 Main Steam Penetration Rooms Heat load from the main steam and feedwater piping traversing through these rooms was evaluated previously.

The inputs and assumptions used in this calculation are very conservative and the small increase in anticipated feedwater process fluid operating temperature

(<2°F) will not have any effect on the calculated or actual overall room temperature.

The calculated heat load in the room is already based on a feedwater design temperature of 4601F, in lieu of lower operating feedwater temperature.

This assessment is applicable and valid for both units.VI.6.3 Turbine Building Heat load from the main steam and feedwater piping in the Turbine Building was evaluated to account for a < 2°F increase in feedwater process fluid temperature.

This evaluation indicates that general Turbine Building area air temperatures may increase by less than a fraction of (0.05'F) a degree. This is reasonable since the minimal increase in the feedwater temperature as compared to all of the other large 75 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION heat loads in the Turbine Building has minimal effect on the Turbine Building air temperature rise. This assessment is applicable and valid for both units.VI.6.4 Auxiliary Feedwater Pump Room and 5' Fan Room Heat load from process piping traversing through these two rooms was previously established.

The less than 2'F increase in feedwater fluid temperature has no effect on the results of the calculation.

A feedwater design temperature of 4607F was used in the analysis; therefore the calculation predicted room temperatures already bound the room conditions expected as a result of the power uprate. This assessment is applicable and valid for both units.VI.6.5 Auxiliary Building There is a minimal amount of piping traversing through the Auxiliary Building to and from the main steam penetration room and the 5' fan room. There is no specific calculation evaluating the heat input from the feedwater piping into this area, given the short run of piping and the minimal increase of feedwater temperature, the effect of air temperature increase in that area is expected to be negligible.

This assessment is applicable and valid for both units.VI.6.6 Control Room Heating, Ventilation, and Air Conditioning System The Control Room (Elevation 45'0") and the Cable Spreading Room (Elevation 27'0") are incorporated into a single year-round air-conditioning system serving the common Control Room for Units 1 and 2.Therefore, the ambient temperature in the Control Room is expected to be the same as the ambient temperature in the Cable Spreading Room. Air handling and refrigeration equipment are redundant.

The Control Room and Cable Spreading Room areas have a third source of cooling, which is not safety-related, in. the form of a water chiller supplying a second set of coils in the safety-related air handling systems.VI.6.7 Auxiliary Building Ventilation System (Auxiliary Building Charcoal Filters)Key parameters for the Auxiliary Building Ventilation System charcoal filters are total flow rates, and total charcoal weights. The charcoal is Barnebey-Cheney

1. 727 (or equivalent) impregnated with 5 weight% iodine compounds.

The flow velocity through the charcoal bed is 40 fpm in all cases and the corresponding residence time is 0.25 seconds. Testing is performed to demonstrate that the installed charcoal absorbers will perform satisfactorily in removing both elemental and organic iodides for design conditions of flow, temperature, and relative humidity.

Periodic testing is conducted to ensure filter efficiencies credited in the accident analysis are maintained.

These key parameters remain unaffected by the MJUR power uprate and, as such, the MUR power uprate has negligible impact on the Auxiliary Building charcoal filters.Based on the above discussions, none of the design, operational or performance requirements of the various area heat removal systems are significantly affected by the slight increase in feedwater fluid temperature.

As such the various HVAC systems are capable to supporting the proposed MUR power uprate.76 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION VII. OTHER VII.1 OPERATOR ACTIONS Operator actions that are part of the Abnormal Operating Procedures (AOP) and Emergency Operating Procedures (EOP) have been reviewed, and it was concluded that the proposed MUR power uprate does not adversely impact the available time for operator actions. The small change in decay heat as a result of the proposed MUR power uprate has a negligible impact on operator response times.VII.2 PROCEDURES, CONTROL ROOM, SIMULATOR, AND TRAINING VII.2.1 Emergency and Abnormal Operating Procedures The EOP and AOP procedures have also been reviewed to assess if there are any impacts to these procedures as a result of the proposed MUR power uprate. The proposed MUR power uprate is being implemented under the administrative controls of Calvert Cliffs design change process. The design change process ensures any impacted procedures are revised prior to the implementation of the power uprate.VII.2.2 Control Room Controls, Displays, and Alarms Section 1.3 describes the physical modifications required to support the implementation of the Caldon LEFM CheckPlus feedwater measurement system. While there are no controls for the LEFM CheckPlus feedwater measurement system located in the Control Room, Control Room Operators have the ability to select the LEFM CheckP~lus system output as the source of input data into the Plant Computer calculation of calorimetric calculation via a control room display interface.

Additionally, the results of the calorimetric calculation are displayed on the Plant Computer to Control Room Operators.

There are no hardwired alarms from the local LEFM CheckPlus System cabinet to the Control Room but system alarms trigger an alarm in the control room annunciation system.Any additional plant hardware modifications potentially required to support the proposed MUR power uprate have been identified.

Also, a review of plant systems has indicated that only minor modifications are necessary (e.g., software modification that redefines the new 100% RTP, rescaling of plant indications to reflect the new 100% RTP). Calvert Cliffs follows the established engineering procedures to ensure the necessary minor modifications are installed prior to implementing the proposed power uprate.VI1.2.3 Control Room Plant Reference Simulator A review of the plant simulator will be conducted, and necessary changes made, prior to implementing the proposed MUR power uprate. The MUR power uprate is being implemented under the administrative controls of-the design change process. As part of this process, any necessary changes to the simulator are identified during the design change review process.VII.2.4 Operator Training Program Prior to actual implementation of the proposed MUR power uprate, training will be conducted to instruct the operations staff on the impact of the uprate on plant operations (e.g., revised scaling for instrumentation, required actions for Caldon LEFM CheckPlus OOS).VII.3 INTENT TO COMPLETE MODIFICATIONS All modifications that are required to support the MUR power uprate will be completed prior to Calvert Cliffs implementing the higher reference thermal power level of 2737 MWt. In addition, Calvert Cliffs 77 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION will ensure all required Operator training in support of this proposed power uprate is completed prior to implementing the higher reference thermal power level.VII.4 TEMPORARY OPERATION ABOVE LICENSED POWER LEVEL Currently Calvert Cliffs uses a rolling eight-hour average of secondary calorimetric power in the surveillance of maximum core power under full-power, steady-state conditions.

Currently, the maximum deviation of the indicated power does not exceed 2754 MWt (102% of 2700 MWt). After the proposed MIR power uprate is implemented, the maximum deviation remains at an upper limit of 2754 MWt (100.6% of 2737 MWt). Additional restrictions on secondary calorimetric power may be implemented in accordance with regulatory guidance separate from this project.VII.5 ENVIRONMENTAL PROTECTION The Environmental Report, the Final Environmental Statement, and supplements to the Environmental Report were reviewed.

The only non-radiological discharge parameter that will be affected by the MUR power uprate is the delta-T across the condenser.

The maximum predicted increase in the delta-T across the condensers after the MUR power uprate is described in Section VI.1.9. It is within the 12'F (max)limit in our discharge permit.The Calvert Cliffsdischarge permit contains the following requirement: "All discharges authorized herein shall be consistent with the terms and conditions of this permit.The discharge of any pollutant identified in this permit at a level in excess of that authorized shall constitute a violation of the terms and condition of this permit. Anticipated facility expansions, production increases or decreases, or process modifications, which will result in new, different, or an increased discharge of pollutants, shall be reported by the permittee by submission of a new application or, if such changes will not violate the effluent limitations specified in this permit, by notice to the Department.

Following such notice, the permit may be modified by the Department to specify and limit any pollutants not previously limited.The MUR power uprate does constitute a production increase that results in a slight increase in the discharge of pollutants, thus Calvert Cliffs will send a letter to Maryland Department of the Environment describing the change that is made and the impact on the effluents.

The delta-T across the condenser is monitored, consistent with our normal practice, during implementation of the MUR power uprate to verify accuracy of the predicted temperature increase.VII.6 10 CFR 51.22 DISCUSSION Title 10 CFR 51.22(c)(9) provides criteria for, and identification of, licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment.

A proposed amendment to an operating license for a facility requires no environmental assessment provided: a) The amendment involves no significant hazards consideration

-This proposed amendment does not involve a significant hazards consideration as previously evaluated in Section 4.3 of Attachment 1.b) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite -The proposed change does not significantly impact installed equipment performance, require significant changes in system operation or significantly increase the release of solid, liquid or gaseous effluents.

The specific activity of the primary and secondary coolant is expected to increase by no more than the percentage increase in power level. Therefore, 78 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION the gaseous and liquid effluent releases are expected to increase from current values by no more than the percentage of increase in power level. Offsite release concentrations and doses continue to be maintained within the limits of 10 CFR Part 20 and 10 CFR Part 50, Appendix I in accordance with the requirements of the Calvert Cliffs ODCM. The proposed change will not result in changes in the operation or design of the gaseous, liquid or solid waste systems, and will not create any new or different radiological release pathways.c) There is no significant increase in individual or cumulative occupational radiation exposure -The proposed change does not cause radiological exposure in excess of the dose criteria for restricted and unrestricted access specified in 10 CFR Part 20. General radiation levels in the plant are expected to increase by no more than the percentage increase in power level. Individual worker exposures will continue to be monitored and be maintained ALARA in accordance with Calvert Cliffs Radiation Protection Program.In summary, the proposed MUR uprate meets the criteria for categorical exclusion from environmental review as identified in 10 CFR 51.22(c)(9) in that the amendment request involves no significant hazards consideration (see Attachment 1), involves no significant change in the types or significant increase in the amount of any effluents that may be released offsite, and involves no significant increase in individual or cumulative occupational radiation exposure.79 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION VIII. CHANGES TO TECHNICAL SPECIFICATIONS.

PROTECTION SYSTEM SETTINGS.AND EMERGENCY SYSTEM SETTINGS INTRODUCTION This section addresses the impact of the proposed change in RTP on Technical Specifications, Protection System Settings, and Emergency System Settings.VIII.1 TECHNICAL SPECIFICATIONS Other than the proposed change to the RTP, there are no other changes required to support the increase in RTP.VIII.2 REACTOR PROTECTIVE SYSTEM The RPS at Calvert Cliffs Units I and 2 includes three trip functions whose settings could be impacted by the increase in the RTP. The three trip functions, as listed in the Technical Specifications Table 3.3.1-1 are:* Power Level -High,* Axial Power Distribution

-High, and* TM/LP.The setpoints/allowable values for the Power Level -High trip are specified in Technical Specifications Table 3.3.1-1. The setpoints/allowable values for the Axial Power Distribution

-High trip and the TM/LP trip are specified in the COLR.The setpoints/allowable values and coefficients for these three trip functions are calculated and/or verified every cycle using the methodology described in References VIII-1, VIII-2, and VIII-3. No changes are required to the methodology as a result of the increase in the RTP. Therefore, the cycle specific calculation and/or verification of the setpoints/allowable values and coefficients for these trip functions appropriately reflect the increase in the RTP. No changes to the Variable High Power Trip setpoints/allowable values in Technical Specification Table 3.3.1-1 or to the Axial Power Distribution or TM/LP trip, settings/allowable values in the COLR have been identified due to the increase in the RTP.VIII.3 LIMITING CONDITIONS FOR OPERATION Four of the LCOs in Technical Specification Section 3.2 could be impacted by the increase in the RTP.These Technical Specification LCOs are:* 3.2.1, Linear Heat Rate (LHR),* 3.2.2, Total Planar Radial Peaking Factor (F.')* 3.2.3, Total Integrated Radial Peaking Factor ( F! ), and* 3.2.5, Axial Shape Index (ASI).The limits for these LCOs are specified in the COLR and are calculated and/or verified every cycle using the methodology described in References VIII-1, VIII-3 and VIII-4. No changes are required to the methodology as a result of the increase in the RTP. Therefore, the cycle specific calculation and/or verification of the limits for these LCOs appropriately reflect the increase in the RTP and the COLR is modified as necessary.

In addition, coefficients for the Better Axial Shape Selection System, which is used to establish the limits for the Axial Shape Index LCO, are updated as necessary each cycle. The cycle specific updates reflect the increase in the RTP.80 ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION VIII.4 EMERGENCY SAFETY FEATURES ACTUATION SYSTEM AND AUXILIARY FEEDWATER ACTUATION SYSTEM The existing Emergency Safety Features Actuation System and AFAS setpoints and response times were used in the justification of the continued applicability of the safety analysis (see Section II). No changes were required or necessary to support the proposed change in RTP.VIII.5 REFERENCES VIII-1 CENPD-199-P, Revision 1-P-A, "C-E Local Power Density and DNB LSSS and LCO Setpoint Methodology for Analog Protection Systems," January 1986 VIII-2 CEN-124(B)-P, Statistical Combination of Uncertainties, Part 1," December 1979 VIII-3 CEN-348(B)-P-A Supplement 1-P-A, "Extended Statistical Combination of Uncertainties," January 1997 VII1-4 CEN-124(B)-P, Statistical Combination of Uncertainties Methodology Part 3, December 1979 81 ENCLOSURE (1)CA06945, Revision 0000, Calorimetric Uncertainty Using the Caldon LEFM CheckPlus Flowmeter Calvert Cliffs Nuclear Power Plant, Inc.August 29, 2008 CA0694),-RAvision 0000 CAXLORiMLI-IC THE LEFM CliECKPLUs FTvMF MEAýUREMENT SYSTEM CALORIMETRIC UNCERTAINTY USING THE LEFM CHECKPLUS FLOW MEASUREMENT SYSTEM: For Calvert Cliffs Nuclear Power Plant UnitslI &2 Calculation No. CCN-!C-080,I Revision 0 Prepared By Hurst Technologies, Corp.Project: CCNAKc Client: Cornstellatioh,.Nuclear

ýcalvert Cliffs Nuclear Power, Plant 1650 CalvertCliffs Parkway Lusby, Maryland 20657-4702 X eu4i P'repared By: 'Checked By: Reviewed By: Approved!

By: Kirk .R. Melso6n Date: 5/22708 Sean M. Matherne 5s/ý2/8 Date, T. H. Crawford R. A. Hunter Date: 5/22,/0.8 5 /02,t 8 Date: PageI OT 30 I.A/CL06 UR( ci CA069,45, Rcvision 0000, CALORIMETRIC UNCERTAINT.Y USING TiE-L'EFM CIIECKPLUSFLOW MEASUREMENT SYSTEM TABLE OFVCONTENTS I. OUMPoSEN..T......S.;N.G....

...........

........ .... ..........

.............

3:2A0 COM PONENT. LIST] NG ... ...... ,....,...... , ..... ,......., ..............

.... .............

3 3.0 M ETH O D O F A NAL YSIS ..................

.. * ....... ....................................................................

.................

o,.,..5 4.0 DESIGN INPUTS ...............................................................

7 5.0 k ASSUM PTIONS .................... , ..................................................................................................................

9 6.0 R EFER ENC ES ..................................................................

........ .......................

................

.11;7.0 CALCULATION

............................

........12...-.........................

8.0 CONCLUSION

.. ... ...... .... .. ... ..............

......,-,....................,..........

.. ..................

Z9 CCN-IC-08001 Rev. 0P Pageeýý00 CA0694'5, R vis'i6 n: 0000 CAIORiMEiRIclt LEFM sU 1.0 PURPOSE The purpose 6f -tIis evaluation is.estimate the uncertainly.in the secondary calorimetric,, as:computed on-le-plant Ciomputer, using the Caldon LEF M CheckPlus ult"asonic flowinca8urcniunt systemi to measure feedwater flow.' Uncctaiiit', is evaluated attheproposicd Appendix.

K tiprated powerof 2737 MW(th).The-Appendix-.

Kp6oWerrepsresefits an 'inreaSe :of approxlinately 1.4%.¢from the current.licecfsed 2700 M W(th).2.0 C'OMPONENPT ISTING,-2.1. Calorimetric-power is-calculated usi ng thefo1llowing, instrumentation:, 2.1.4. ,Feedwat.e:IFlow 2!1. L1L. Feedwawer:fbw~ismeasiired using the (CThdon'L-EMCheeckPlis ultrasoric flow-measurementi system.,Principle comrponents of the LEFM CheckPlus system consist f:.a. Metering'Section

-`The metering sectionis:a:spoolý-pice~i nstalled.iin the feedwvater header toýeach steam generator.

The meteringsection-c~nsists o.f8itransducer pairs.E'uipment I'Ds b. Electronic.Unit -The electronicuni't-(one per plant)'sequencesthe-operation 0fthe transducers and calculates volumctric flow,-temperature, 'and mass-flow.

Th~e,,igitalvoutput of the electronic unit provides input to the plant computer via the plant data network.EqtpmenitIDs (Based on AMAG Cabinet:Numbers

-ToB.eTVer dL"ea)i CPU ICi209- Unit Cal don LEFM CheckPlus

-2CPU2C209-Ufiit.2: CeckPlis .Ele6irdfiic Unit 2.1.2. 'Feedwater Pressure.2.1.2.)..

Feed'water pressure is measured from-pressuretransmitters Trorm taps installed in the'meterihg section of the flowmeasurement system.Equipmcnt 1Ds-1 b tninied,-2.1.2-2.

The feedwater pressure transmitters provide~input to-the flow measurement system electronic unit, which transmits tihejinformation to-the.plant computer viazthe plantcidata :2.1.2.3.

Feedwatr presstfrei s used for, a. Calculation offeedwater mass flow.b: Calculation odf feedxvater enthalpy,., 12.1.3. Feedwatce Temnperature 211. 3A- ! F 6edwatcr temperature is measured from RTDs. nstal ed" in ..the fecdwatcr header to eac hsteam generator.

Equipment lDs 11TE4516 7 1 4SGOFeed~vater InleýfTemperature:

K1 TE'51,7 -12 SO :Fed?.wftei lInlet 2TE451 6:--21, SG-Feedwater Inlet Temperature SO-Feedwater-InietTemperature.

2.1:3.21 The f6ed;.ater fenipzratu're-RTDs

'pro'dide ihnp'utjithe plaiit.

2.1):J.. Feedwatertemperatureýis used.for-the The feedwaterARTbDsare independent of the temperature measurementused.to calct!late feedwater flow.CCN] IC-. 08001:l Rev: 0 Page:3I0f 30

'A06945, Revisioni.O0000 Y UN-I-¥ iNG.THELEEM CHECkPLIS FLOW.M:FXSURIFýMENT SY'ST-EM 2.1 .4. Main Steam Pressure I2.A.4. Main. steampressure-is measured-.from pressure transmitters installed~in t efsteam, hcaders downstream .of each steam generator.

Equipment DsýTi"l3991 -II Main StiamiHeader l'resure IPT4008.-

,12 ;MhinStedafiarHeader Piessuire 2PT3991 ,-21 Main. Steam .Header Pressure 2PT4008,-%22 Main' Steamf Header Pressure 2.1.4A2. Main SiedamPressure is used~fOrfthe alc'u'liioj,ofstee6m enthalpy.,2.1.5 Steam Generator ,B,!lowdown FlowSteam generator blowdown tlow is determined fiom indicated total ,Equipnent, iDs, IFT4089 W- Unit I BD Tank-Effluent Flow 2FT40891ý-

Unit.2 BD Tank Effluent Flow 2.1:5.2. Blowdown flow for, each steam generator is not measuired directlyibut'is'tabnually-input to the.plant computer inýaccordance'with Reference.6.7.

The blowdown'flow input represents~the total blowdown flow from each steam generator.

2.1.6: Calorimetric Constants Calorimetric constants are assigned to calorimetric inputs not directIly measured',by.plant insirument's such as W Heat addition to the reactor ,coolant system{(RCS),from thetpressurizer

ýheatetsTand reactor coolalnt'pumps,*: Net hieati 0ssfror6'thC RCS,.t letdowh flow.o. Net-heat, lo,-s- frbm, the.RCS thfroughbinsulation.

• Steam generato exit steaim quality.CdiCN-lC0800l Rcv. 0lg"o'3 CA06945,Revision 0000 C.ALOKIMTIC UNCI RINTY iNS THE L;kEFM CHECKPLUS F:LOW, MEASURMENT'SYSITEM,`3.0 ME O -I O- ANAIL YSIS-3J.-

nuses'ti tfieithdddlo'y

'established.t.in ES-028"' rin'truin Loopii : Unceraifty aiidSetp6int Metiodology" (Refertence6.1).'

3.2,. Sign Conlvention:

3.2.1. Oncertainties areapplied to actor values; as opposed to calculatcd~or indicatcd values.3.22. Un'certainty is positive when the calc -lted&valueis.greaiterthan the: actua`l:value.

Uncertainty.

is.n~gative.when the indicated or calciulitedývahe is less than the actual lvalue.32:3; calculated is less than actua powerandthe plant is being operated with indibated power near the rated thermal power limit, thie Appendix K thermial power -limit i'ma'y be, exceeded., "There fore, :onlyýthe negativecomponent of, cal6rimetiricuncertainty needs, tobe evaludted.

3.2.4. Since only negative calorimetricuncertainly is considered, the meihodology established in Reference 6:11, Sect ion :Correction fo-r Setpoints w itha Single Side of Interest" may :be:appli ed 3,3. The secondary calorimetric:cornputation is performed, using various inputl process pafameters, such as.FeedwateriFlow;Sleam-Presstire,.Biowdown Flow, etc., which, are either measured or estimated.

The Uncertainty .in ihe measurementjor estimate of each of these process parameters is, analyzedto determihe its individual impact on the secondary calorimetric c6mputation.

These impacts are, then combinedtob dt'termrinie an overall uincertainty in the secondary calorimetric computation.

)3.1. Thefimpactof`the uncertainty fora process parameter is analyzed for a given condition (actual, process-value). .he.potential.

range ofnominal-conditions is determined from analysis ofhistorical plant readings, as listed ard summarized in Reference 6.13. The limiting conditions are iased on pooled data from Unit I and Unit 2, whiichhare different because ofdiIfferences in secndnry plant design.

theidat'ais used to provide conservative overgallimnisits tO applyJo both Ubnits.Reference,6,13 extrgpolates.theinteasuredýupper and low6r range values f6rthe current 100%po6Weir limiits to the new', po'weruprate condition s;- based on iobservati ois0f the chanige-,in the paranmeter with inpow'er.

The final output-of Referencc,6..1i s.a set of process limitis,.

wich-arL;e osidered."'indicated,,yaues.`

3.'3.2. The sensitivity ofthe ec'ohdary calorimetric comptitation;.isýssessed-to detcrfiine whether an upper or lower pr6cess valueiss more cohiservativýe for-use in the s~eondair' et.orimetric uncertainty analysis., In some casesthis is difficult to assess computing~the effects, since a given.pararncternmay impact sevcral aspects of thc sccondary, caloriimctric computation. (For instance,:the, proccss valuc for fee dwater temperature affects-the impact of-the feedwater flow, Teedwater pressure and fecdwvatcr tcmpcraturc uncertainties, whichzare:

all used in the secondary calorimetric, computation.)

In these theoverall effects are numericall assessed to determine the most cons erv ati\ ei i nit for use;ý Jrhemioust conserv"ati-v liniit is used in Uthefinal uncertainty anialysis., 3.3.1 For each process parameter, a directional error of an inpUt process parameter readingwVilil produce a corresponding directional error, in the-secondary calorimetric computation.

For theesecondary calorimetric unccrtaintx analys~is, only thosc producing'a liowcr-than-ac tual power c&mpuitktion will be consi&redc siriceia 1ower-ithah-actiial-.

pover comrputafion Would, cause at4 ahighti--reatcLurpuv ei-.lcv. Thercfurý;

imfip of each uncertaity in theinputprucess -wllbeonsidered to determine the -pptopi riatce direction for tincerta -intycboisideration.-

3.3.A. The impact of instruiment unicertainty-for a 21iven process parameter is determnined by a few, simp5le sitep. First,'ihe contribution of the input parameter to the secondary calorimetric coiputation is determined from the nomifinil condition (actual process the coitriboti6n is reý-ý, comnputed, witlh erroir applied to the aýct4al, proccss VIlue.(iindicatedvalue).

Thl,-diffcrence iih thesetý6 comptutations is the imrnact of the uncerta inty of the individual process parameter o the secondary calor imetric. com9putatio~n., 3.3.5.: Finial uncertainty`

imphcts fromn ieach<~-t~he input 'o'e~ss parameters ,are comibined ttodete'rine.

an,:oNI r .aunIcertaIintyin thie. secoiudary.

Calorjimitetric comnsputation.

CCN-iC-08o01 Rv., 0a page:5,of 3 ,0 CA069451.Revision 0000 CALORIMEIT'RIC UNCER-TAINTY USI'NG THE LEFM ICHECKPIKS FLOW MEASUREMENT SYSTEM 3A4. The calculated uncdrtaifiny is-a'boundingguricertiinty, applicable to both 6nits. atthe proposed Appendix.K uprated:pover-bf2737 N4W(th). Thebourndifig uncertainty is based upoi!i conservative:values' for fceddNater-floW uincertainty.

Tlhe-final niargin'value',s exjie'ssed as :aipercent bfthefiew uprated power ievel o;f 273.7 MW(th).3.5. The ASME 1967 steam tables aie the cuirent~basis fortli plaiihconiputer deternilnation of theimodyfiamic propel-tiesl.

3.6: The computations~performed were calculated tfoseverai significant digits. Hand veri'cfiotin utilizing the r " oundqed valuicsi'1iay."result;in sligitly differenit-reSulis due to round offe,-rrors.

-3.7. No0computer-codes were ýutilized in' the perfformance ofthis calculation..

31.8. Unless otheirWise noted, U ,il be used to designate random unceartainti'es',and'-B-will be us'ed to desigfiate bias uncertainties.

31.9., Subý.dipts: (M), -Maximum'Value of1 Input Parameter (M1)- Additional Maximum VaIue of Input Parameter for-Analysis Only (Notan ýActuaiiLimiit OhlyUsed for Feedwater Temperaturein thissAnalysis)(m) M inimunum Value of Input Parameter (SS) -Sinigle Side of Interest ACT -Actual Value-Calorimetric Po'vcr with No UnccrtaintiesApplicd

-Uncertainty Using.Actual Value for Parameters

-,Actual Va lue of Paraumeter BD -Blowdown BDT -Total Blowdown CAL -Calorimetric.

CALC --

FS.- Satuirated Liqfuid F W,- Feedwater GS -Saturated Viapor-hpc -Saturated'V~por'Enthalpy

-Saturated.

Liquid-Enthalpy h ,s. -Iat.Pratled Liq i.i i dEnth alpy, hiw,- Feedwater Enthalpy hos -SaturatiedVapor Enthalpy IND,- Indicated Values.INPUT.- Evaluation of Calorimetric Poweri drUncertainty

ýith!,Uncertaihty Applied.to a Selected Iniput MBD -Blo6wd6wn MassFlow'

Rate:-MBDT -Total:Blovvdon Mass Flow Rate, MFW -Feedwater Mass Flo\wRate NEI -NetCalbrimetric Uncertainty Net Contributibn to0uncertainty from each input OTHER-- Contributionýto6Cabimetriic Power fro, Other Inputs to-the Seconday Ca lorimetric Cazlculation, not based'on ters PC -Plant Como,'Ut&PFW -

PSTM -Steam Pressuire SG -Steam Gereratot6r SG1 -First Header Steam Generator SG2-Sccond -eader-Steam Generator STM -Steam TFW -Feedwajer Temperaturer CC-N-IC-08001 Rev. 0P'P~ge 6.of3,0 CA06945.,Revisioir 0000 CALORIMJEfRIC UNCERTAINrTYUSING THE LEEM CiECKP-usTLow MEASUIREMENTI SYSiEM C40 DESIGN INPUTS 4.J. EQUATIONS FOR CA.ELORlMETRlC-POWER:

4AA.!. Per,.Rfercnice 6.2, tlke gross thermal outlut'of dne. steam generator is computedfromth hcxpression:

Qso ..(..v .. ..)h.+.hs-hrwI+MnD(hrs-hpw).where;MFpwis feeddivaei:

now, MB 5) is blo'wdhWn fl6w,;hFrs is the'fluifd coimpoinent of steami enthalpy, hs, 0 is~thc Vapr~corinp6nent of steam ienthaply, hoW isithe feedwater enthalpy (comprcssed liquidd),.and Xis, the steam quality.4.1.2. For, a steam quality, of I' (no moisturel carryover), the 0abov eexpr essi6n isimplifies to QsG XhGS." hvW)+;(MBDXhrs

-ýhS')Qso MWhS hw-Mor~4.1.3. Calorimetric Constant 4:1.3.1. To determnine reactor power fr6m steamri are made to accouinltfdr'heat"trnitigfer to/fromr the Reactor. Coolarit System from sources other, than the reactor and sinks other..than the steam'generators, such asthe lheat added by the pressurizer heaters,and heat losses ,through pipe.insulation.

4.1.3.2. These corrections areinpt as -onstarits6t -the caloifimetriccalcdulti6n.

4.1.3.3. For conveniencethis calculaifon represents the-net adjusitrentas a4single cons.ta.nt, QOTH:ER.4.1.4. Calorimetric power-is the sum of the gross thermal:output of.both:stcam gcncrators-and the calorimetric~constant.

QCAI= Qs62-- Q-TJHER-4'.,2. INPUT UNCERTAINTi ES: 4.2.1. Input unreirtdiritis foir themriieasieedrparamete-rs, Iwithexception of feedwater.

pressure(See Assumipti ohs)ýire Su mmari zed in the table below:ýParafijeter Raiiddhi."U, P,6sitivei Bia'SB.B Temperature, Tvw,. deg. ýF'Steam Pressute;Pjý,ýIPSl 91- 19.80 +3.40(Total BlowdowniFlow, M m, klbm/hr 4ý2:22

References:

4.2.2.1. Main Steam' Pressure-;_,Reference 6.3, 4.22.2. Eecd'gti"rTe per-the R'66rence6.5' 4.23. B -lowd6wn.Flow.

4.2.131. Total bfowýdown flow uncertaintylis evaluatedlin Referencew6.4,as a function.ofindicated blowdown flow and power.4.23.2; From a review ofReference64,.

blowdo~vntfow at50 gpm indicated flow'ounds the uncertainty at higher blowdown!

flowvrates and wilIlbe.used for this evaluation.

CCN-1 C-08001 Rev. 0 Pagek7 oCf30, CA06945, Revision 0000 CAL[OIRIMIEIRIC[

U USING TlE:L.EFMHCl-iEKPtItJS FLbOW MEASUREMiENiT SYSTEM U.2;33.. 'Since ifite sign f6r uncertaiiitiesdin this ref6rence-is inot-clear, the largesthimagnitido.'6f tcficetaiity,(p6sitive or negative) for indicated fliW ,il be u'sed fiiid bothi'.the'-pbsitivde and: 'negative'dire'cti6nS.

,412.3.4.

The-uncertainty ofReference 6.4iS expressed ih units of-klbm/hr aId roundreedtup'to.

the~nearest U'I kibm/h rf6i'rsimplicit)y, and conservatisim.i 421..,E PlantC.tompuier.C:al.culation of Enthapies:

From Reference.6"2,.the'unccrtainty-of the plant computer calculation,of enthalpy is Up PC +6. 010, BT.U /Ibm 412.5. SteamQuality A42.5.1. Per Reference 6.8,steim quality , X, is set'. t 1,1 As, shown belowthiis is, awconseivatiie input and-uncertainties are not considered..

Ii if carryoveris considered;, he gr6ss therinal 6utput .f.onec teamn generator is.represeatcddby

~Mw l~ -X~FS+ XhGs;ý- hFI (MDXFS G$)w¢hich reducesi to sc MFWA'hFS -hrw+XX r AllXh -h~s b. Sincequality only has values:of I orless, the calculated .thermal-output,using a quality-of 1;will:alwaysbe equial to or greater than actual tmal output .-c. This results -iiaposibve uncertainty and does-not need, obe considered per Section 3.2.4:246. CialofirmetrieC onstant Biases:-4.2'6:1. The biases of the calorihietric-constantsused in Reference-6.6 were reviewedby Reference 6"10..As sign conventnionis.f .the:biases.waýreverified i6'b& cofsisient-wiihSection,-32.Corfectionis tW iaseswere rmde by:Referen ce 6.6.1..4.2.6.2: :Per the negatiie .v~ue of biases needs tobe 4'.2;63. If thd th6e biaLses oftlte,ýc6nstdantt'are.differeht:between/,Unit I And:Unit, 2, is usedto "evaluate

ýcalori metric uncertainty.

4.2.6.41 The biases: asociaited "xýiththe`

c' const~ahts are"siimmamrized!below.

CONSTANT BIAS UNITS PAK0021, Pressurizer.-ieater Input 0 MW PAK0026.'Reactor Coo0lan!tPump Heat Addition -0.6I MW PAK0022. RCSHIeat Loss-. -0.35 MW PAK0024A Letdbwn FlbowHeat Loss -3.19 MBTU/hr CCN-IC-08001 Rev.O 0 Pagc:,18,of,3o' CAO 6945 Revision 0000 CALORIMETkIC UNCERTXN ' USING .Il_ EFMMCiEcKPLUS FLOWMEASUtREMENT SYSTEM 5.0O ASSUMPTIONS, A .. Feedwater Floý: 5.1A. Feedwater-'flow ufri{ertainty is detefminffed b the vendor, Cameronfinternational Corporation'(formally Caldoh, Inclrporated) hased upon.hydraulicrmodeling and testingpat-.Ahydraulic lahrartory (typicallyv Aiden Lab,). Although-,,t eI ing has, riotbeen comýleted, typica -uncert inties are ssThan '04%.5:1.2; Fofrthis 6ealuatiofii, a174/se'ati.Je-uncrtainty of+/- 0.5000% adtuaI.flo\,vMwll be 6sed.* THIS ASSUMPT, iON WILL BE VERIFIED UP'ON RECEIPIT O'F-THE VEND6OR: UNCERTAINTY FOR FEEDWATER FLOW.'.5.2. Feedwater.

Pressure 5.21. The Caldon topical report,'Ref 6.12,-and tIfe supplement to. the topical report,,Ref.

6,12.1, assume a.pressur6euncertaintyof+/-

15.00 PSI" which wiMllbe used in this evaluation.

• THIS ASSUMPTION WILL BE VERIFIED UPON SELECTION 6F THE PRESSURE INSTRUMENTATION A ND EVALUATION OF THE INSTALLATION.

5: 2-2.Actual. feedwater pressures attthe entrance to the steam generatorsare hot kinoWnr. Calorim-etric uncertaintY ill bee valuatedusing steam:genemator pressure This results'in a-conservative calculation o ffccdwatcr cnihalpy sinceactual prcssurc'at the inlet to ihe steam generator m ust:be greaterthan steam generator pressure. .see 71.21for furither discussion.

' THISASSUMPTION DOES NOT., REQUIRE VERIFICATION.

THE SELECTED-PRESSURE CAN CONTINUE TO BE USED-WITHOUT KNOWING ACTUAL FEEDWA-TER PRESSURE..

5.2.3. The saine .feedwater*.pres'siare-instiuthientation:is usd, foribdtlifeedwitei" enthalpy andfeedwater flow.This assumes the p'riniciple cb'ntributiorioto.:feedwater'flový is density. F6r agiVentemperature, wahigher than actual pressure increaes.s'derisity',-resultingiin a higher feedwater flow, measurement.

ýFeedwater'enthalpy.,alsoincreases, but theincrease isirinimal.

Fr m 6inspection'`

nf-thesteam generator tlhermalo,0utput calculation, in'feedwater fl measurement resUlts in a ,higher than actualcalorimetric power computation.

An increase, ihn eedwater enthalpy results in a lower than actual calorimetric power computation.

Therefore4 the~effects are offsetting,.and'it-is more conscrvativc'io trcat the uncertaintics in icedwdacr flowdd wntltcalpr as independent..

• THIS ASSUMPTIONWILLBE VERIFIED UPON RECEIPT OF THE VENDOR-UNCERTAINTY FOR FEED WATERIFLOW.

5.3. Bo16wdown' FlowDistribution 5.31'. Thc' bloWdown, flow mcasurcmcnt.uscd'inthc plant calorimctric is total blowdown.

flow.5.3.2. From. References 6:7.I'and 6.7.2. total indicated blowdown flow through, both steam generators is.limited to 180 gpm. Maximum indicated blowdown flowthrough-asingle steam generator is 1550,gpm (Unitý 1). TheUnit 2 procedure does- notprovide a, single steam generatorI limit. Therefore-the Unit .Iimit from Reference6.7.l is used.r5.33. This ,activ ity evaluates, bloxdo,:vn.assumingithe' fo6llowing:

distribution o f-blowdown ,flow: 5..3.3,,1. "Total'bls.vd6wuivflb-h wissictaf t th 'maimum- pernissible te (1 80 GPM-'t ,25.6 klbmlhr,,per Referened'6.7).

BIowd6whiflow'for 6ne'steafin gefineator is 'st ýtthe maximum per1insible flow:rate,(150 GPM= 107Tklbn/hir,, per Reference 6.7).5.3.3.2. Total:bl1down.

flown is set a t the maximum prniissible flot..,rite (180 OPM =1[2$.6 klbmlhr, per Rcferifce,,6;7)..

Bl6wVdd.Vn floW is evenly distributed-between the ,team- geheiradors.

' THIS ASS'0MPTION DOb'ES NOT REQUIREVERIFICATION'S1NCE IT, IS FURTHER EVALUATEDb FORCONSERVATSM -WITHIN THI-1S.CALCI.!LATION.

CtCN-IC-'09001 R~ex'. 0:ae9 f3 Plage 99.f 30.

(CA06945.

Revision.'0000 CAIL-ORIýMETRiCUNCEiRTINT'Y USING T1IE)LEFM CIiE'CCKPLUSFLOW MEASURMENt SYSTEMI 5.4., Assumed Planht-Paf-rheters at,2737 MW , THIS ASSUMPTIO3N'WIVLL-B E. VERIFiE'D UPONiPOW'tER ESCNLATION,.ALTHOUGH SLIGHTýCH.AN"G"ES IN PO}WER.SHO)ULD,4T SIGNIF ICANTLY.AFFECTiCALORIMETRIC UINCERT'AINTY.

,54.A.. For parameters other than blowdowrn flow and feedwater press ute,-conservative valucsi based upon trendsin plan tlparameters following a reactor startuprafter a refuelingoutage are-selected to-maximize calorimetr ic uncertainty. "T, he summary values and the data used in the-determ nation-are included wifliif! ,1 fe ren~ct 6'.'3.52.412. The foil lowing the, bounding~condiiions fon- each .parameter'.

indicated vahji,. with excepfionof~blwoydo~wn:flw; as~proyideda from Reference6.1

3. The-maximum.

indiated values have been'rounded up to-the nearest whole number. The minimum indicated values have:been rounded'to thenearest whole number.,'M ifnimum. Ma1ximum Param; feter (IndicatedcValue) (hddicated.Value).

Feedw'ateriflow, -Fw, 5932 6,78 klbmn/hr FeedWiterTempera'ture, 432, 443 432-.443 Feedwater Pressure, P.? ,,8 PSIA .m85r4ssur_, _ 76_,_A_8____4 Steam-plressure, Ps- ,IS'SA 89 4 CCNAC-O-8001 Rev. 0.lage I Oof 30 C CtAo6Q6945, Revision 00010 CALORIMETRIC&UNCERTAINtY

'USING THE LE FMýC46ckpLus FLOW MEASUREMEi4T.SYsIfEm 6.1. ES-=28, Instrument Loop6Uncertaint Ad&Setp0int'MethodologyRevisionI 6.2. VTM 12f389-249,9Cont 06,l Speic Phiýt ýConipiter, R:evision 25" 6.3. IYCALCI-93-037,Uncertainty'Cflcualaion For The Plant Computer Indicati onO'f Main Siearn'Pressure,'Revision 1 6.4. DCALC CA045.64, UncertaintyCalculatiion for thelBlowdoWn Flow Input to:the Secondary.

HeatBalance, Revision 0.6.5. DCALC CA00470,.LoOpUncertainly For FecdwaterRTDS.

Revision 0'6.6.

UnceraintiesOf The Secondary Calorime.ficConstants'F'r UnitIl 2; Revision.0 6.6.A. CCN i O9Q3-07201)!,ARevision 0 6.7. O0eratingdPr"scedures 67J.1. oi-068A-1, Blo~wdownSystem.

Revision 39 (,Urnit' 1)6.7.,2. OQ8A-'2, Biowdown System, Revision 3 T (Uniit 2)6.8. SP 094, SyVterf 094-Setpoiht File, Rvi.i6n.-9 6.9. ASME Steam Tables, FifthEdition

(.1967ASME Steam Tables)6.10. ESP ES200400492-060, Review/Revise Calorimetric Constafits, Revisionr 0 6:.11. ISA-RP67.04.02-2000..Methbdol6oi es: b ithe Determinaii6n-6f Setoointi tsrt'Nuclear Safety Related lnstritmentation, .1/112000 612. idon, Inc. Enginring Report-80P; opical Report. Improvinh Therma :Power Accuracy and Plant Safety While itncreasmings6prating Power-LevelUsin6gihe Check Syst.m, Revision 5,.October2001 6.12.1. Caldon,,inc.

Engineering Report-I 57P. Suj)oIlement to jTonicat Report ER-AOP Basis for a Power Uprate°WiWh theLEFM. Check or LEFM CheciPlus'Svstem.

Revi siionm #,2arch, 199.7 6.13. .DMLS.#1DEO7881 .D. A. Dvorak to Filfe, Dated'March 31.2008,,"Estimated"Parameters for:Calorimetric'Power for a %. Ao pendix.K Uprate"-CCN-I-0,08001 Rev. 0a pagelf, ,o:cf 30, CA0694.5, Revision 0000 CALORIMETRIcUNCERTAINTY USING THELEFM.iCHECKPFuus FLOW MEASUREMENT SYSTEM 7.0 CALCULATION

71. EVAILUAXT'Id OF UNlCERTAINTY 7.1:1. The calculation of calorimetric uncertainty:lhas three mfnajorcompon'ents, QGI, Qso2ý and.QODmEtt.

Calorimetric:unceraintymay bhcevaluatedby eyaluating the'uncertainty'for.-eadh mopronent, then statistically combining the results.7.1.2.

6feach~majoib icomprised of'indiVidual compbtuefits.

For example, Qso, is co mprised offeed fldw;,blowdown floW;,feedWater-enthaiWy (determined:rfriom feedwvater pressure andttemperature inputs), and steam enthalpies (d:cetermined by stehm ;pressure input). The;totapluncertainty is determined by evaluating the uncertainty for each component, then statistically combining-the results.7.1.3. :for thegrOss thermadloutput of a,:steam generatorI Qso, the~contributiofot~r each input 1to, uncertainty is determined.

from USGINPUT = (QSG)ACT ,,;iere (Q9S )CALC-INPUT i"s the gross thermal output.of the steam generator determined by varying the, selected input:parameter by its uncertainty while usihg actual values for the other inputs, and (QsG)ACT is the gross thermal output ofthe steam generator determined by using actual values forall inputs.7: 1.4. Biases are similarly determined where:eac'h input parameter.is~varied by it associated bias.1.2. SELECTED FEEDWATER:PRESSURE AND TEMPERAT'uRE 7.2.1. Feedwater temperature an-d pressure are used to calculatefeecwater-enthalpy.

'Feedwater pressure is also used& nithe dlcu lat ion0of fedwater flow, but the, effect o ffeedwater:,pressureon feedwater flow is included in the uncertainty 9of the measured flow rate. (See Assumption-5.2-3..)

7.2:2 Enthalpies*are takenfrom Reference76;9.

Feedwater enthalpies for the range of interestateshowti.summariz.edbelow:.hw" BTU/lIbm _____P,,,~A

_____Trrw, DEGF 800 850 900 .950 ý1(000, 420 397:35 397M40 397A45 397.50 397.55: 430 408.29 408.33 408.37. 408.41 408.46 440 419.31 419:35 419,38 419'.42 419.45 450 430.43 430'46 430.49 430.52' .430.55: 46,0 441.66 44 1.68 441.70 .441-.,72 441.74 7.2.2.1L.

Feedwater Pressure a. Enthalpy increases as pressure increases.

From inspec.tion ofthe.expression for steam.generatorgross thermal outputý a negative calorimetric.uncertainty-willIresult if indicated pressure than-actual pregsure., Therefore, a posiiiveuncertainty.

valtiewill be applicd for- ugeiri the secondary calorimnetric unlcertainity analysis.b. Maxiniizingithe difference between steam enthalpy and feedwvaterenthalpyrvil i resulItinihe greatest calorimetric uncertainty-contribution from feedwater flowv. The differencecis-maximized by minimizing f6edwater enihalpy, using lower values for feedwater pressure: c. relatike.

changePin.feedwater enthal'py with'tempcrature.decreases:as pressure increas~esý Therefore, lower valules ofprvssure maximize theconitfibution 0ffeedwater'temperaturesto uncertainty-.CCN-IC'-0'800'.

Rev... 02 Pap '11of30' CA06945, RevisioWO0000 LEFM EFLOW:MEASUREMENT SYSTEM dA For% given temperature, therelativechange in enthalpywith s e a esentially constant.

of-feedxyatepre's'surecto'.unrc-ertainty,'d`oesrinot yaryl Sulbstantially withechanges i iniiaiii pressure: e. In sumimari, lower'valies 0flressure maximize theicohtributifiS, of feed,,ater floW.and: tempetature'euncertainties t6:.cdilorimetric unricertainty .Pr6cess~pressufe values.do'not significantly impact the contribution-'f feedwater pressure uncertainties to caloriinetric uncertainty.

f. Calorimefric dirncertairity islmaxii-zized byaiissuiming minimum feedwater pribssure with indicated pressure greater thafi'actual iressue, (PIW.IND(,,)>

PFW.ACT(m))-

72.2.2. 'The actual value.offeedwater pressure.is determined by subtracting mneasurementiuncertIainty frdri.ithe ninimurm indicated valueo'f feedwvai, er pres'suer

..PFW-AC'Tm)

=FW-INDtm)

-UpFW S , PFw. A r(f.,). Reference PSLA, P. PSIA" -Secti6ns, 854 1:.00 839.00 5A..2; 5.2.1 7.2. 2; 3: Feedwater Tiemperature

a. Enthalpy increases as temperature increases.

From inspection of the expressibn for steam generato'r gross thermal output, a nriegative-:calorimetricuncertainty wilt result.if-indicated:

tempera ture-is.greater:than actualtemfipefature..

Thusi a negative uincertainty'yalue wilL'be appliedifor.

se:1in the secondary calorimetric uncertainty analysis.b. Maximizing the 'difference between .teaminenttialpy ahdifeeder ethialpy willresult in.the greatest'cailrimetfic:'uncertainty contributi6fifrom feedwater.

flowe. The'difference is: maxinmized

'by mninim izing feedwater enthalpy, using lower'.valucs for'fecdwaterdtemperatuie.

c; The relativeehange in feedwater enthlaipy with pfressure decreases astemperiatureriiicreases.

Therefore, lower 'valueso of emper~atu'r e maximizethe conttr~ibution oif-tedwatei rsue Unceriaintý.

1. Fora given pressure, the relative change in:enthalpy increases as temperaiureincreases.

Therefore, higher values of temperature-maximize the contribution of feedwateetemperature to uncertainty.

e.. in summay,,.lowcervalucsoftemperature maximizethe contribution of feedwater flow and maeedwaiertpressure to calotrinetric unctertainty while, higherivalues of feedwat"ertemperature maximiize ,the corttribttto of feedwater' temnperatturelto uncertainy.

CCN-'IC-0800 I R:eV. 0 Page 13,'of-30' CA06945, Revisioi, 0000 CALORIMEaI:C.

UNCERTAiNTFY'USIN6 TiIE LEFMT CHEckPLtJS FLOW MEASURLrEMEN-r Sys'Ij;M f. Calor metri,.unceti-tainty will be evaluadted at bothl-iinimumtand~maxinum feedwater temPeratdres, %ith indiiated-temnperaturfe greaiter than.actualVternperature

(.TFwIN,(,)ý>

TFw'T(, )andT'w.1i')>

TF.wAtý,I)., In addition, analysis is perf6rmedat ,a higher tenperature (454 deg. F)ithan :the those~specified~within.Reference 6.1 3

>. TFs'Aý'(Kii))

to confirm the errorttrend; an.thus be confident of choosin.g the most con servative process conditionmfor analysis.Tfw.INhm)

D( I TPnw (m)' Reference deg.. F ...... Cdeg. F Sections 432 1,. 88: '430.12 5.4.12 A42.1[ T¢v¢.Ac.(M~)

UTPF ()N Refe'rence

_deg. F deg.: F Sections 443 1m88. 441i.-12 ý5.4.2, 42. 1,

' TPjV.4NDQý11)

Refefince deg.. F .... .deg,; F. Sections,:45 1.88 452.12 .4. 2.1 7.3. SELECTED.STEAM PRESSURE 7.3.1. Steampressure isused to calculate,steam~enthaipy.

Entihalpies are taken from Re ference 6.9. Steam enihalpies forthe range of interest areshown summarized below:-:h sý hJG,, P,,M,.PSIA.

BTU/Ibrn BTU/Ibm 750ý 1200.7 699.8, 760 .1200;4. 6977 770 1200.2: 695'7.780 .ii99.9 693.6, 79,0 1,199.7 69 V6 800 1199.94 689'.6 810 1"1599-11 687.6 820.. 1:1.98.8 685'5, 830) 119W8.5 683.55 840: 1198.2' 681L5 850' 1198.0 679.5 860 1197.7. 677.6 870 11097.3 -675.6: 880 w 197A 67316 890. 1`96:71 .671.6'7.132.. Sincethe miss flow rate. offeedWatcr is§substantially" reaterthan ti"e mass w rate through a.s ingl e~steamgenerator. (the ratio~of feedvater-flow isi> 65):.the'prinoipaA effect on calorimetric:uncertaintyis ihe:contribution to calorimet'ric'uncertainty.

from :teedwaterflovw.

Maximiiing~thc diff&rence between steam enthalpy and feed enthalpy willrcsuilti'ithCegreatest calorimetric urtceriainiy contriibution from.

flow. T.he .differenceis.maxirnized by maximizing the saturated vapor enthalpy; Therefore,,a lower steamr pressure maximizes the coht.ribution of feedw(atier flow to-calorimefric uncerttainty.

CCN-IC-*8001 Rev..0a o14'-oftT3 0

CA0694h, Revision 0000 CALORIMErI-X1C UNCERTAINTY USING THE-LEEM.CHECKPLUSFLOW MEASuREME?4 SYSTEM" 7.3.3. Steam Pressure Cdfitribtition to. Uncertainty 7:33..1. The siaturatedvapdr enthalpy, hGs; decreaseslas pressure:inreases.

From inspectionof the lex.prcisidni foi stcani,-gcn&'rator gross calo'rimetric unce6rifity*willfrsult frointhm e chahgem.in-h 0 s:if ihdicated pressurie i than actual, piressure..

73.3.2. The differeinc6ebetween the satutatedvapor

' and saturaied iiquidenthalpies, hFG; also decreases as SFrom inspection of the kxrps~ orsan erior gro's 'sý herina].ou tout,a-teqgative calorimeftricudn'certainty

'vill resultfro m-'t'he ch'an~ge in hFo if ýindicnted kressure is lessr.than 'actual=

7:3.33: Referring to ihej-information withiin able-of Sectibn7.3:1A for 75 0 a / nd .760 PSIA, for al 0 psi increasein ihndickted presture above-ae;thi pressure,-.h 0 5 s hanges-by approximiately

-0.'3BTUWIbm, Whie-higqchanges by Apipr6oimately

--2-.1 .BTU/1ifim (a:pprxiiatiel' 7 timesgrerter thfiarnthe

change'iifi h 0 s).7.3.3.4. For calorimetric uncertainty, the change in'hos is amplified by the feedwater flow raie while the chafigein is aiiplified'by the'blowdbwn 7.3Y.3.5, Since the mass flow ratepoff eedwater issubstantially'igreater than the:mass flow rate'f blowdown tthrough ausin gle steuiivgenerator,.an increase in indicutedup ressureis a net--negati've contribution to calorimetric uncertainty...

7.3.4. The-charigein h 0 s and.the'difference betweenrthe saturated vapor and-saturated liquid enthalpies, h-G, are essentially consýtant as steam-pressure chlanges.

Therefore, a.change in initial pressure has a: miniimal effect on calorimetric uncertiintyi.

7:35. Caibrimetfictidncertainty,.is maximized byiassulming imini mum steam pressure With iridicated predsure'greater than iactual,,-pressure.(Ps.i~MINIm>

D PfNs4MAC Tm).The actuial valh' fsteami.pressureris~de femined b"s'-btractihg.

mearurehent Ui certaihty and positive bias from the minimum indicated value-of steam pressure.pSTM-ACT(ni):, = Iwl() --UPSTB -BPSTM Y.SrM.630iND(rTs BPSNI P .StM ACT(m)j Ref~erenice PSIA, I ...... PSIA Sections.819 19.80 3.40 795.80 5.4.2- 4.2.1-NOTE,- The NTIST.steam

tables show that .the change-jin-the enthalpy

increases as pressure increases, w.,hilethe change in the difference between satUrated'vaporýenthalpy'an.d saturated, liquid enthalpy decreases, buthe:4 hangersis-light.

This*4res ults ino.pposite calorimetric-uncertain-'fro"nsteam'-pressuireuhicirainty.

Since the eiass-fl6Wvrateoffeedwaterflow is greatertliun themiassflow rate of blowdown.

the contributi6n tocalorimetric uncertainty fromsteam pressure-uncertainty Will be;greatei at htigher steami pressures.

However, the~overall-contribution tocalorimetric uncertainty from .feedwaterfow-is also much .greater than the contiibutionfrom.

team.pressure.

Therefore, maximizing hs;-by'using lower values of steam generator pressure,-

results in a conservative assessment of calorimetric uncertainty.

73.6. For this eva*luation,.calorinetric-uncertainty will beýevaluated at mini mum-steam pressureto .maximize steamenthalpy.

CCN-IC;-08001 Rev. 0--Pag I lý,5of 3o0 CA06945, Revision 0000 C AL ORI M-ERTRIC::U NC ERTAPlJTY EFM, CHEckPIUS FLOW [iMEASUtit-MINT SvSTEM 7.4. ENTHALPIES USED IN EVALUATION 71.4.1. Feedwater Enthaipy Enthalpies used-in this evaluation are, surmnrarized in the table below. Feedwatef ehthalpiesare derived from interpolation of:values in R.eference,6.9BT/IJbm @ T and P 'Pressure, PSIA pIrw.ACT(,, )Terperatute;, dg. F 839 854 TFw-AWfi i} 430.w12 408.4534 :408.4654_1__:_____,__

432.00 410:5252"F___:__T__

441.12 420.5858 9 420,5967 TFW.rND(M) 443.00 422.6749 :-:. : : A TM,-^cMl_

.452:,,2 432:.8325 432.8409_T_..V.N(M t * '454.00 434N9423 .,.," 7.4.2. Steam Eiihalpy 714..1. Since.steam pressure has anapolicable bias; enthapli3esare notivaoluat'ed at indicated pressure.Entbihlpies 4ire aply ingrandom ndbia c6fomponefis actunipresiire:.

7.4:2.2. Sieam enthalpies used i nthis evaluation~are::summarizend iil>the ,aIe be low. Enthalpies are derivedl:from interpolation of values in Reference 6.9.Pressure, PSIA hGs, BTUbrM hbýe,, BTt'lbm.I 1199.5260 690.44,0Q PS.ACT,&m)+B PTMI 799.20 1[199.4240 689.7600 P Ar UPSTtM ..... 81,5,.60 11'98;9320 06864240 7.5. FEEDW ATER.FLOW UNýCERTAN.TY-CONTRIBUTIONt T75..1. A.negative results fr6rmianeg(itivicfeedwater floýmeasuremnent uncertaifity.,.(Indicated flowv<Actualflow).

Also, th&.contributionto calorimetriýc uncertainty from feedwater flow-'measurement~isma.'imized by maximizing feedwater flow. Thcrefore, the maximum indicated-flow is used in the:evaluatiorno~fcalorimetric uncertainty.

7:5.2. The ma.kimum-actual flow. isideteriiained froim the- maximum ihdicated fl6w ard, the 'associated Orinceilaiity.

MFWIND(M)

= MFV-WACT(M):(i -UFW)'-. T M(-UFIW)D(M IýMi; I WINJ) .U .MFvwAýýAtM')

.ýRfcrincc

.klbm/hr .klbm/hr .Sectis: 6178 0:5000% .6209.05 -5.l"2, 4.2f1 7.5231. For eachszieam generator, t[(MFwXhGs -hFw) (MBDXhFG:)]CALC-CCN-IC=08001 Rev..O.Page 1'6 of30 CA06945:'Rvvisiun 0000 USING THE LEFM -CHECPIMUS'FLOW MEASUREMENT SYSTEM 7.5'3.12 Foranh~rroi flow the expression iedtJce§ tb USG.MFW = (MFwIND)(hGS -hFW.)-*(FWý-ACTXhtGS hew) or-USG-MFW -(MFWýIND MFW-.ACTAXhýGS-hFW).'7.5.T33. eir'ris ihe sme for-each steamgenemator, USIG.-MFW,/

USG1-1F.W

=UsG2=MFW.

7.5.4. Contri.butioqn, tqoalrimeTricr Uncertainty, 7.5.,4l. Using the expfessi6fi above, the; contributionfiffeedwater fldoý, uncertainty to the;gross thermal.outpuit of, on ne steam, geneait'oreis

.:i. ,Minimumn lndicated iFeedwater Tei~i~eratt~re

b. Maximum Indicated Feedwater'lTe.nmperature I h-.-w, BTU/Ibm D M) pw.-". .... hGs7,BTU/JbM USGoM.W Reference klbm/hr klbm/hr @PSmI-AC-r(,)

MBTU/hr Sections" " "PRV-ACTý(ni) 6178.00 6209.05 420.5858 -7.5.2,7.422, 7.4.1-3,1 ;05ý -"7,78.9402

-2.12..5.27 .Z 7-41 c. Additional Maximum Feedwater Temperature (Beyond Upper Limit)MFWNDM)' M .WAtYl{NI

..hs,:~q~b BF,.Tu/lbm' MwIdbmihr Mklbn hr @o-;-AGs, m USG.MFW Reference..... kl.bm..hr-Wsf..ACT :TM BTUIJ/hr Sections,... ..__ _ __ PFW-ACT(mi)

,,:6178.00 6209.05' 11995260 :432.8325

-23.:82 75? 742;7.4.1-31,05_ 766.6935 4. The tables above ,demonstratethat the contributiton of feed\waiterfloWto calorim'etric incertninty i; mAximi7ed uising Iowervues of fe~edwater, temperattre.

7:5A4.2. T1henrietcontributionito caloriinetricfireidertaifity is UCAL-MFW xU SG=-MFW FW -Temperature UCAIL4IFW,_NBTU/hr@TW.ACT(rn)

-.-3'73117.@TFw,.,,______

-34.1990 I @T.wAC(M1)

-33;6614 CCNAC-08001 Rev. 0 p~ageý 7 of30 CA06945, Revision.0000 CALORiMEiTRIk:UNCER'TAIN'[

Y USING.THE LEFM CHECKPLUS FL5w MFASUREMEN'T SýSTEM'7.6. BLOWDOWN FLOW

'CONTRIBUTION:

7.6.1 : 'General ons for. Uncertai nty: Since the same boun~ding conidilionqsare estbi.bished'8or main steam :pressure,.f}edwater-temperature:and.

feedwater pressure, the net thermal output ofboihi steam generatqrsQsj I+'QsG_4,. canrbe re-written as,+SG Q =b ý'2M , Xh,&ý --h hFw)+ 'MBOTXFS h~s wvhere MioTr is ,the total blowdown flow.The, contributioti to caloriimetric.,tincertainty from total blbwdown flow measurementiS UCAL-MBDT

'=kSP -I-G .Q:SG2.,+

QOTHER )cALC: -MBDT' A(QSG1 +,Q SG2 -+QOTHER.)ACT-MBbT C'AL.Mnr { XhGS _ -,h

_A vuXhF6)+ Q.THER]t(M )v s ~ -P (MBDT .AC~F)+/-QOTHEk]

UCAL-MBDT

.

7.6.2; From inspectibn of the abov, e:equations, a' negativecial6fifietri.unc:ertaiiity

ýresiults.fr6rm

at ipositivd total blowdown flow measurement uncertainty

'since hFG ispositive, (Indicated flow> Actual flo.)7.6.3. The~net con'tribufi.on toclorimetrii U'ncertainty fr6m blowdown is'.hFG,'BTU/lbm UCAL.MBDT, Reference.

UMBbT, klbm/hr @PSTM.ACThii.

MBTU/hr Sections 7-9 690.4400 -5.4545 4,2.1 7.7. FEEDWATERENTHALPY UNCERTAINTY 717. 1. Fee dwtei.Teniperat ufe 7.7.1.1. General Equatibns for Uncertainty:

a.. For each steam generator, USG-TFW { -wh~rVoohn3~Tw

b. For'an error in feedwater ternperaturemeasurement;,theexpression reduces'to USGTFWy FW XhGs -hFW-.CALC ) ( MF Xh~s-hwAT orr USG.TW = (~ XhFACT -hFw..cATc where hýw.AcT.is ,evaluatedat actual feedwater temperature and pressure while'hEw.I4D is eValhatcd at, indiibated

'dedwater iemperature and actual pressure.c.Theerroris-thc~saie for each stearnmgenera~tr,:

USG-TFW = = USG2wTFW CCN-IC-08001 RevW0a Pdge 18 of 3a cA06945 .Revision 9000 NERTAINTY.USINGTHiL EFTM CHEC4KPLUS F MF.XSUREMENt.SgS.Ti:M 7.37..2., C6otribuiion to"Calorimetric uncerainty

a. Usiihg the'expressi on above,.the contribution of feedwater temperature,'measurernent uncertainty to the gross thermal output ofonesteam generaioris

(.1) Miniimuml ididated Feedw'atlef Temperatu re hF.WACT: hFW.CALC MFvACT(M), BTU/Ibm BTU/Ibm ..Gw,. 'Referencee klbmihr @Wi:zc'r(,*

@Thw b-,,~),. MBTU/hr Sections______________

FW-ACT(,6nY Pi W-ACTflu)

________ ________*6209.05 408.4534 410.5252, _12.863,7 7.5.2,,7.4.1 O620'965 2.607 1'8'(2) Makximum lndibated.Feedwkater'Termperature hm.-ACT hiw-CALcBTU/lbm. BTU!Ibm UsaTw, Reference klbm/hr @TFw.ACT(M), @Twirt.D), MBTU/hr Sections, PW-ANCT 1 T) ,WACTWnM)i6209.05 24:M585$.

422.6749.9

-119713 7:5.2i 7.4.'6209.05 -2.0891 1 2.(3,) Additiinal Maximum 'EcedwaterTemperature. (Beyond Upper.Limit):

MN FW:.A(M), BTU/IbAf BTU/lbm Reference klbmhr T @TýACT(M)

@TF'NO(M MBTUOhr 'Section's

____________

PFW2ACTýrn)

~ F.-A- m) ________ _______6209.05 432.8325 '434'.9423: "j30997 7.5.2, 7.!6209.'05 .12,J098 .......97_ 7-5.2? 7.4....(4) The tables abOve demonstrate:

that the contribution o'ffeedWater temperature measurement to calorimetricuncertainty is maximized-tsing igher values o ffeedwater.temperature.

b. Ti'e net contribution to lcalorimetricb;uncertainty is U.CA.-TFW

-x G-xFW FF 1 ICAL-TFW, FW Temiperature v Uh MBTU/hr.

-181920@TFW-AC(M)

-18.3442@jpw-Acm,',;i ) -1:8.5257 77T2.,

7.72.1. General Equations for Uncertainty:

a.' "For each steam generator,{[(MFw.Xh GS-, hFW -(M BD XhFGI)CALC

-PFW,.USG..PFW -l'[(M~Fzw:X.GS

-hFýw)- (MBD'F'G)]ACT J GcN-C-O800Iff)rtcv,.0.

Pgc19of3 pg.g *19.. o3 f) U.

CA06945;:RuV'ii6ni.0000 CALORiMRIpiC THE LEM CHECKPLUS FLOWMEASU RrEMrENT SYSTEM b. Ror*anefror infeedw,,atet preýssbue.measufem"nttihe cxpression reduces t6 USG-PFW = MFW XhGS I-_hFW-CALC

)-. W XGS, -F W'A.CT ), o r*USG-PFW, = (MF~WXhFw-AcT -hFW-CALC) where-hFW-,Ac is evaluated at:actual'feedwaterttemperatUre-and pressure,while hFw:iND is, evaluated at actual feedwater temperature and indictated pressure: "c; Thze crror is the same for each steýAmigeerator, U.SGPFW ='USG1-PFW

= USG2-PFW 7.7:2.2. Contribuiion.to Calorimetric Uncertainty-a. Using the express ionaboye, the contributionof feedwater pressure measurement uncertainty tothegrosssthermal output of onesteam generator is (1) Mifiinmum hidicated:Feedwater.

Temperature hNV.ACT hFW CALC'Mi.w-AC-frtN BTU/lbm BTU/Ibm Ud46p-w Reference klbm/hr @TW,'."A&mi)'

@TFw.ACT(m), MBTU/hr Sections ,P FW.ACT(M;' "PF, WND~n)6209:05 408A453A 0.0744 T.5.2,7.4L 65209:0'5

,00;' -0.0.12744______

,7-:5.2,'7:4.

I_(2) Maximuimffindicated F1eedWiter Tempeatui:e 4WAT h C LC'mBTU/b n UsGPFw Reference klbim/hr @TRv.acr(M), @Tw.A&r(M), MBTU/hf Sections 6209.05 420.5858 420.5967 '-00680 7".2; 7:4"1 6209.-05 -. 0_______________

(3) Additional Maximum Feedwaterjemperature 11 rW.CAL.U MFW-Axc-r(Nit BTU/Ibm BTU/Ibm Uso.Prw Reference kIlbnmihr

@TrW.ACr(M

), @TOW.ACT(NMI)!, MBTU/hr Sections_______________~~~~~

'~wA~~ p~ND(n, ________6209.05 432:8325 432M8409 -0.0519 7.5.2, 7.4.1 6209'.05 -0.0084 .9___ 1 7.5' .2, 7.4....(4ý) T'he tables a'inovedenmonstrmje.hat the contribution of feedwater pressurenmeasurement.'to

ýcalorimetiric uncertainty is maximizediusing lower values of feedwater temlperature.

b. The net contribution to calorimetric uncertaintyis UtALýPFI=Jid 3 G FV Temperature MU-hr MBTU/hr ,@TFW-ACr(,.

70.1053'0T1w.AC"quN

-0 07.3@T~W.CTMO

-0U734 CCNMIC-0800t IRev. 0e Pageý2O'of 3oQ CA06945, Revisioii 0000 CALORIMETRICUNCERTAINTY USING THE LEFM CHECKPL.JS FLOW'MEASUREMENT SYSTEM 7.8. MAIN. STEAM ENTHALPY' UNCERTAINTY FROMIPRESSURE MEASUREMENT 7:81. RAnd6m Compoihent of Pressure Measurement:-

7.8. 1,. 1. General Equationsfor Uncertainty:

a-, FofeachWsteam ,generator.

h~w -M 8 0)(F: )CALC.- PSTM!.{Mý-%[(MFW-XhGS -hFw.)-(MBDA1hFG)IACT

b. For-anerrorin steam pressure measurement.,-thieexprcssion reduces to

{ M .....XhGSCALC) (M 0 3 X ..FG-CALC' t 6r YSG-PTM l(MFW I GS-ACTI)A MBD)(F 'Or)I'USGPST = [(MW Xh hs

CALC -h HFG ACT'r)]1 wNhere hGs.ACT adnd hfG-ACT are evaluated.ata'ctual steam-, pressure, whil :hos.cXLc and hFoCALC are evaluated by applýing the ratidomcompoient of steam pressure mneasturementi uncertainty to4he:actual pressure.,c. vary be'iween-steam-geriirai6rs;,th~e eror may f6r, eich genfeiator.

7.8.1.2. Contribution to:Calorimetric Unceriainty

a. Usihg'theexpressibn aýve",-the contributioniof steam pressure measurementuincertainty'to thegrbss ili rmal;au.tpuo(of one-steamT generator is (I) Case. -Maximum :Flow -'hrough OneSteam Oenerat6rMaxirnurfiTotal Flow hGS-CALC ,hFcCALC MFW.AiT0M)

BcTUIIbn .h os.Ai--A-hvo .USC!PSTNI... .... BTrilPT klbnlAr @(PsrM-_Acrtm)

+ :BTU1IbmT klbm/hr @(PSTM-ACT(m)

BU/lbm MBTU/hr 1 4 PSTM) @PsTN1ArtL UpsTM) @PsrM-Cc!

,)6209.05: 1198.9320 1199.5260 10.7.00 686.424o0 690A4400z 6209.05 w-0,5940 107.00' ,-4.0,160 -Y.32585_.3688;,1,7

-429:71 1" 6209:0c5s 1198.9320 1199.5260 1'8.60 686.42401 690A400-6209:05 -0.5940 , ,1860 -4.0160 -3.6135-74.70 Note 1: Values Obtained from Sections 7.4:2, 7.5.2 and,5.3.3.

Note 2: The. table above issplit -into two setsof three rows-each.

The:secondr6owin each set provides a-difference in the enthatlpies as compuled in row I, and the thiid row.0pro0ides the flow multiplied-by .he differece in" cnthalpies.,;CCN-I:C-_0800 r Rev. 0: Page 21 of 30, CA06945;,Revisiuh-0000 CALR .TRI'c;UNCER:AINTY¥U'SING THqELEFMKCHHEKPI:US Fýow MEkSU REMENT SYSTEM (2), :Case 2- Exenly DistributedBlowvdoevniFlows, Maximuam Total FIowý, LCh6s.Acr r.C hFG.-ACT MVW..zik.

BTU/ibm hK.e M a, BTUiAl'bm, m0~......l BTU/lbm B BTU/lbmW/bm klbmUhr

+ klbm/hr @(Psr.Cm'+

'MBTU/Ih" UASTr) rTm),UPSTM)

M3TM'r.6209;05 A 198.9320 1 1995260 62ý86 686.4240 69.0.4400 6209.05 -0.59040 62.'80. -4.0160 -3'43'60-3688.17 -252.20.'6209.05o ] 198.9320 I1'9905260, 62:80 686.4240.

690.4400 62169.05 -0:5940 62.80 -4.06160 -3J4360-3688.17 -252.20 A Note: Valuds Obtained ffom+.Sections 7-4.2, 7,5.2 and'5.3.3.

Note 2: The table above'is splitinto twqosetsof three rows each. The second row in eacqhsetprovides a differentce' in the enthalpies-asscomputed in row,, and thejthird row provides the flow multiplied by the -difference in enthalpies.

b;- The net contributionrioical6rimetiic uncertainty is URS ALP2: +'(uSG2PSTM) 2 cA'S ,.Usol-srM SG2;PSTM 'UCALolntm C MBTU/hr ,MBTU/hi I MBTU/hr I -3.2585 7-3.6135 -4`8657 2' ,-3 .4360 -3.4360 .-'4i8592 This-table shows'that the contributiongn of steam-pressure measurement to calorimetlric uncertainty is maximized by assuming maximum flow through one steam generator ,and maximum t6talflow (Case 1), but the effect is slight.7.:8.2. Bias'component of pressure, measurement:

7.8.2.1. General Equations for Uncertainty:

a. For~each-sieam generator,.

f [(MFWVhGSh3 w) (8 O#XFG )iICALC -TM}BSG.PFW l-(MFWXhdb -hFw-(Mbbo.Xhpc,,)IC b For'an errnr in ..4tcamprcssiire n uefisirenentrthe expressi6ivredui~esio BSG-PSTMi

= ..F : XG .CALC. -.(M 6 .oFGCALC)_r IN F-[(MF hSACt,)- (MwXXh BSG-PSTM = [(MFWXhGSCALC hGS- ACTr ) -XhF h FG-ACT )1 where h6s.(,cT.and'hFb-ACT are eyaluated at actual, steam-pressure., while hcs.'CALC and hFG-CALC are evaluate~dby applying.

ihe ebias com:ponent ofsiearnm'pressure measurement uncertainiy'to the actuaL pressure: c. As Idemonstrated .preuioqs.i', thkecontributionof steam pressur. measurement to calorimetric uncertintv i's maximized by assuniIng maximum flopv-through one:steam:generatorpand:is.the OnIly"ase evaluated when~e.aluating bias.CCNIC-08001 Rev. 0 Page 3220f 30:

CA06945, Revision 0000 CALýORI!MrERiC UNCERtAINTY USING TiiPLEFM CHEcKPLUs MEýSUREMEN:T SYSTErM 7.8'22. C6ntriibuti6io.t6oCal6rimetric Bi'as ,a., Using the exNpression above, the contribution of steam 'pressure mecasuremen.t bias. to the gross therrjal bOdtpul uojlie 01ain generator is hos~cxLC FOCLhr.AT S.............

................

BTUI/lbm SG PSTM UPSTTS ~ !~ 2CIT(M) -'5N1ACTOýmIT*I. 'UP~sTr, ... ...6209.05 1199.4240 1199.5260.

107.00. 689.7600 690.4400, 6209:05 4.01020 107.00. *0.6800 -0.5606,"633.3 2 -77'2:.76M 6209,05, 1199.'4240 11990.5260 1.i60 689.7600W 690.4400 6209.05 .... 18.60 .6800 -0.627.-633.32 12'... ... ..i 65: Note: :.Values Obtaincd from Sctions 74l2,7.5.2 and 5.3..Notel2':The table aboy'e"is, split nito.two sets of three rf6ws each The' sieofidxrow in each set provides :a:differenice in the enthalpies'as c'omputedfin rowlhandthe third roW;provides the flowmultiplied by the enthalpies,.

b. fihenet contribution to calorimetric biasAis B.CA'L iSTM = B SG'i-P!STI:M" B ~2-PSTrM BsG.psTTm BSG2-PSTM BCALPSTM ,MBTU/hr:I MBTU/hi MBTU/hr 4056066 -0.6207. 4.11812 7.;83. Th ilnet n6ntribution of~stcamn pressure inesurement crror to winceirai'ity is the sumnof tlhe Fandofni contributinto

uncertaity .and thebias contfibution to uncertainty.

ýUCAL;FTM MBTi'hr BCLPTM~MBTLJ/hr-4:8657 -1.1812 " -6k0469, 7.9. PLANT COM PUT E R-UNCE"RTAINýTY CON'TRIBUTION' 7.9.l, Feedwater-Enthalpy 7.9,1.1 GCeneýral Equitions

]forUincertainty y:'a. For each steam generator.

[(MFw XhGS -hrw,) -(MDOb FG)AAcT b., 'An errori in the co.mputaii6n 6f inthalpy i's deterriiined, from UsGPC(hFW)

= (MFW XhFW.ACT -hFWCALC)or

ýUSG...PC(hF.W)

=(MFW XuP 0)CCNAlC-08001 Reev. 0 CA06945-.

Rv,isiui:

00660 CALORIMEkICVUNCERTAlNY( USINU TIELEFM CHF, CKPLusýFLOW'M .SIA fýM ENY SYS tE:M c. The e"rfi is the.saine for each steamn genhrf1or, Us -.PC.(hWV,) ,-.C(hfEW U -GCCbFW,)a'id~the eict coniribution ,tbcalorirn~tric uncertainly is., 719.:2. Using,the expression above I the-cortribution of feedwater,.enitha.1py.un.eirtainty to the ealorimetric un certainty from the p:lan't comP-ute"r c6mputation o f.enihaipy is IMFIIwýAr(m) 1Uc USG.c.(hrw)

UCA Reference klbm/nr h l 3TtJ'lbm V MBTu/hr IMBTU/hr Sections'6209.05 -0.10, -0.6209 750.8178 .5.2 4.2.4 7.91:2. Main Steam Enthalpy, Saturated'Vapor 7.9.2.1. Theplant computer does not calculated hNdAirectly but takes the difference between-the,'calculatedfsaturated'vapor enthalpy and the saturated hiquidentihalpy.

7.9.2.2. GenieralEquations for UncertaintS,:

a. For each steam generator,{RMFW*hGS -hFpw)-0MBp~hGS

--h FS)]AL- Pit Us,'PCOGS

-Xh ., (s -h=s )h cl b. An error in the computatioro m0iaiin'steamsaturted;vaporenfthalpy'is deterlmined fromn-. w (- BD.XcXhG Lc} .or

-(MFW 7 MBeDXhGS-CALC

-hGS._ACT) or_ = .M -MDXU )c. Si 6icl-b 6lWdowih iflow can'vary bet we-n stniirgenerators, the error may not be thetesamici for reach generator.

Asdemonstrated preý,iiously,.a'ssuminrg niaximumrbl'wdowln flow %through.onesteam generatoris the:more:conservatiVe approacti.

The net contribution to'calorimetric uncertainty is UCAL .PC(hGS), =.VUS(jSGPC(hGS) ) A' (USG2pc(hGS)Y) 7.92.3. Using the, exprcssion rabove ýthe conttibuiti'o"ndf ifairnsteat saturated.

vapot~enthaip ,ouncertainty

'to c UnhcrtaiuntyfIrum fitIiC pldILitiilputer cuirputatuloni feithulpy is*MFW-ACT(m), M61 UPC UsG.PC(hGS) .UcAL.pc(tds)

Referehce klbm/hr' klblmhr B1I1U/Ibm MBTU/hr MBTUlhr Sections 6209.05 107.0 ý0'ý 10 -0;6102:6102.0 0 .10 1 _ -0.8692 7.5.2, 53, 4 6209.05 ,18.6 _ 0, 010 6190.4, 1 0-0690 7.9.3. Main Stearii Enthalpy,,Satuirated Liquid, 7.9.3. 1: The :plant computer does notcalculate hFG yire ;y bqut.takes the difference between the:calculated ,aturated vapor enihalpy-and the SaturatedI liQIuid enhalpy..CCN-IC-OttO9 lRev. 0 tPýg6624 6f,30 CA06945. Revisioni 6000'CArORI.EITRIcUNCERTA Nr,.Yi USING THE LEFM-C-fEckPL0S F£ow SYSTEM 7.9'32'. General Equations for Uncaity: a For, eachsteam*ngenerator,=[(FWVXIIGS 7hýFW:)-MBDoXhGS

-hFS")ICALC

-PC ,SG, .-[(MFwXhGs-hFW)-(MXhG GS-hFS)]AcT

b. An e~rrour ii'inte e~uillput'dtior ufiintej autd liquid'entialpy isideterfii~i~fied6661 UsG-PC(SF)

= (MBD XhFS-ACT -h'FS.CALC) or USGr.PC(hFS)

= (MBDXUPC)c. Since,.blowdown flow can vary between steam generators, the error may not bethe, same for each generator"?

As dem9onstrated previoutsly, a'ssuming makimum blwdown flow through one steam fgefieiriitodis the~moreýcoriservative apprfoach.

ULJCAL-PC(hFS)

= j(1USG,_PC(hFS),2÷+.(UsG2_PC(hFS)2 7.93.3,3 Using'the expression-above, the, bntribuition;bf,,rainýsteam saturated liquid enthalpy uncertainty to the ca!orimetricpuneertainty from the plant computer computation of enthalpy isýM13a d ,,s-PC(hFs), UcAL.PC(hpS).

Refernce:;klbill/hf BTU/lbrii.

MBTU/ir MBTU/hir, 'Scctions 107A0 -0.n10 0.0107 1 18.6-0.0 j -0.019 -0.0109 5.3,4-2.4 i8;6. _. 0A.l0 -'0.001.9: " ... " '7:9A4. The' c6ffibined tinenrainty for tfhplarit&:

mputer m aleulatioii

&fenthalpies isgiehby'J'(cA-Pchr~v))2~ (JCL.~PCt~G) 2 +(YAL-rcqiirsi)y'

[UCAIL-PC(hJF-W)

UCLr'Q~ CAL.Pt,(hrS)

UCAL-PC 1 BTU/'hr ýBTU/hrý BTU/hr .MBTU/hrJ-0.87S8l -0,.8692 7 0.0109, -L2356-J 7.10. CALORIMETRIC, CONSTANT BIAS 7;10,1. Since the inputs to QOTIIER only consist of bias terms, thecontribution of Qo-nER to net-calorimetric u nceltainty is the~sumof all biases associated vith the inputs tO'QOTHER.

7.10.2. A conversion of 3.41 21 &IMBTU/hr/MW is used when summing the biases. Thetbtalibias,.CA... IcOTIIF.R, is expressed inlunits of MB'T U/hr. rounded up to the nearest 0;0001 MiBTU/hr.REFERENCE CONSTANT NEGATIVE ,BIAS 'UNITS SECTIONS PAK002 I, Ptessurizer.leite'.

Input 0 MW 412.6.4 pAK9026,.

Reactor, Coolant Pump -eat Addition -0.61 MW 4,2.6.4 PAK0022;,RCS Iheat Loss -0.35 MW 4.216A4 PAK0024; Letdown Flow Heat Loss! -,3. i9 MBiU/hr 4126.4'BAL;OTInpR

-'6A4657 MBITU/hr CCN -1 C-08001 Rev; 0 Page:25 :of 3 0 CA06945, Revisipn 0000 CALORIMETRIC UNCERTAINTY USING THE LEFM CHECKPLUS FLOW.MEASUREMENT SYSTEM 7.1 1.- NET CALORI:M ETFRIC. RAN:DOM. UNCER:TAINTY 7.11.1. Thecontributi6n:of all random terms to, ret cali'rimetric.,uncertaiity is determihed-using.thelmfiost.

limiting uncaert ainties inpuit.7:1.1.1 .1.Feedwater-Te:miperatiuie:

a. Feedwaier temperature was evaluated at:minimum and maximum actual temperatures, as.well as,.a temperature

ýhigher than the actual temperature range. toconfirm theltrend n, the ,,uncertainty data. Feedlwater temperaturesli~ripac( feecrflo'measurementuncertainty, feedwater temiperatur-e measuretefirtunierainty.

and eed'ater preIssure In Iaisuremrnt.

uncertaintly.

The fihal termperature to use in the assessmenrtn6f caloiriictric incertiinty is determined by using the most conservaitivedimnit:

I UcALM~, J UCAL.TFW, 'u CAPFW, I ss ,F*W Temperature

! CA1L;MRV "icLTW UAýLP~W SP-RSS" r MBTU/hr ; MBtUihr j MBTU/hr _ _ "__________

I S4; 7317 j -18.1920'

'70-*1053 39.2078,.@TW;ACT(A) j -34:::1990'

-`18.3442 j --0.0962 38.8084@rATnj -33.6614 9 .0I.557 7 -0.0734 3'8.4226*b. "Itis-observed that.theovera IHcalorimetric uncertaintiy increases with decreasing feedwater temperature in a-near: iinearfashion over the temperaturerainge of interest.

Therefore, the min'imum'alun of em limiing assessment of calorimetric uncertainty.

7.11.1.2.,Blowdown Flow -As demonstrated previously, calorimetric uncertainty is maximized by maximizing blowdoWn flow through one steam :generator.

7.14.2. General Equation: UCAL =-,(UcAL-MMF

)"+ (UCALMBDT) 2 + (UcA..TFW) 2 + (UCAL.PFW) 2 + (UCALSTM)

+.(UcALpC):.MRTtJ/hr M3TtJ/hr MlBTiJlhr M3Tf(Jhr MlRTr J/hr MBrTIJ/hr MRTI.J/hr 34.17317 -5.4545 -18"1920 A-0.I53 .4.8657 -1,2356 -39.9024 7. 11.3. Single Side of lnterest Cot0rectibn 7.11 13 IS.Since only-negative calorimetrlcuncenainty is considered, thie methodology, established ihn Reference 6.11, Section,8.1i,"'Correct ionforSetpoints withla Single Side of Interest"7, may. be apPlied.7.] .3.2--he calorimetric uncertainty calculated previously is based upon a 95% confidence level (1.96 standard deviations)

The 'random component ofcalorimetric.uncertainty.'mayabe-reducedby.a correction of (I 645/I.96).

1.645 UCAv-(sS)

= ' ,(UAL) = -33A4895:MBTUfhr, 1.96 7.1-2. CALORIMETRIC. ,BIASS 7M12.1. Calorim'etric bias is thesum 6f.allYbias cohnpoinefts.

7.12.2. Contribufions.to calorimetric bias'are limited to the calorimetric constants.and main steam pressure.RAL-OijIIR, RCALlPSThIr, RCAL MBTU/hr MB'FU/.hr MBTU/hr ,6A4657 1.112 -T:6469 CC"N-C-08001 Rev. P.Pageý6 f01"0 CA06945, Revisiun 0000 SCALORIMEETRICUNCERTAIl'IY USING THE LEFM' CECKPLUS FLOW MEASUREIit&T.SYýSTEM 7.+3. NET CALORIMETRIC UNCERTAINTYNet d norimeiric uncertainty is UCALNET -UCAU.L BCAL 7. A'.2. ,convcrs'onoqfJ:4!

2141 M13BTU/hr/MW'is used to, express uncertainty in MW._7.13:3. Uncertaihty ise,xpressedlin

%RTP by dividing the, in MW. by 2737 MW.7. 3.4. Net Uncertainhty f UCAI.ý BCAL U.-,Al.-P4T fUCAI.-N~i U4i;11 MBTU/hr ,MBTUI/hr MBTU/hr MW J %RTP ,-39.9O2'

-7.6469 -'17.5493

-13.9353 -0.5091%7..13.5. Net Uncerta(inty, Single SJide of Interest BCAL 1 UCAL.NCT J C 3 AL.NET(SS,)

UA.Et MBTU/hr MBTU/hr JMBTU/hr.

MW %RTP-33--4895

-7.6469 -41.f365 , --120559 U .0.4405%7.14. MARGIN-7.14.1. Available

margin is obtained.by

adding thenetuncetfhinty,,tolthe Appendix:K iower"lfmitc6f2754 MWzand:subtracting the&rated thermal.power.of 2737, MW.7.14'.2. Marginhis expressed i dividing 2737 MW.7. 14.'3.J Available.Margih

', Rated.Appendix K CARtedma LUx 1 i4, Margin Margin IThermalrI.P Limit Power MW .. MW %RIP 2754 -2737 -13,9351 '310647 01 120%7,14A4. Available'Margihý.

Single Side of Inteiest Rated Apendix K Thermal, UALET(SS)

Marginý Margin.Limit' o MW MW %RTP ,Powe&2754 2737 -12.0559 4.9441 0:.1806%7.15. RELATIVE CONTRIBUTIONS OF EACH INPUT TO OVERALL UNCERTAINTY Procviding a'relative contributionfi permhits methbd of combiniing random atid bias compbnientsiof for each nput t'6eva uatethe effect' f eachtinput on the:bveraII'calofiiiietric uncertainty., ,7.15.1; T'he relative, contributlon of each random inpu.tto calorimetric.

uncertainty is givenby U' INP UUT. / -ipr') .7.15.2. This expression can be combined with'theassociated bias input.to find thenet contributiori of thatýinput.to calorimetric uncertai-nty CCN4IC08001 Rev.0e Pagie 127"of.30 CA069.45, Rcvision 0000 CALORIMETRIC UNCERTAINTYN

'USING THE LEFM CHECKPIUS FLOW MEASURýEMENTISYSTEM 7.15.3., The net'contribution of-each inputotb net calorimietfic uncertainty can theni be foufid'b, U imU, E, =INPUT-N U CAL-NET and is summarized inhthb follbwing table: 'IiU'UT I I" ". " '" (U ý" 1 j'-Nil)UA NPUTr'Nra-y,'

INPUT _ MBTU/hr MBTU/hr, MBTU/hr MBTU/hr FeedwaterFlbw

-34,7317 -30,2310 -30.2310 63.578%Blowdtown Flow -5A545 -0.7456 0:75 -0.7 156 1.56%Feedwater Temperaturc

-18.1920 .,8.2939 '4' , 8.M293 17.443%Feedwater Pressure '0:i053 -0.0003 7. 0.0003 ..[ ." 0.00 1%Steaim Pressureý

-48657 -0.5933 11812 1.7746, '. 33732%Pilant.CompuierýComputatio 0 of Enthalpies

-1.2356 -0.0383 70,0383 0.080%Other Inputs _-64657 _ -6.46571 3.598%Totals ,3939024 -76a469 -47.5493 100.000%71 5.4. SiigleeSi.de'0ofinfe'resi:

7.15.4.1.

A sinifilar appr&oach is used to find the rcIativc~c-ontfibution of cacti :input tb calorimetric for' the sirfgl side of ineresi.7.15.4.2.

For sirigle side of interest, the relative contribution of'each input to calorihiti~ic' uhcertainty is, gi'en by UINPUT(SS):

1.65 (UINPT UNPUT 1.96 UCAL.UINPUI"NET(SS) andN U, INPUT- NET(SS)U INPLT,-NET(SS)%N ITM 121%= .CAL...NET(ss) 7.15.4.3.

The relative contri butions to calorimetric uncertainty),using-single side oflihte-est is summarized in ,!theciable.bel6w

UINPUT (UINPm.Tss))

'1 3 INUT (UINPUT-NCT(SS)),INPUT MBTi/hr MBTU/hr MBTU/hr. ' MB3TU/hr*Feedwater FlOw -29.ý1498

-25.3725 ,7,, -25.3725 6j,.675%Blowdown Flow -4657797 : 0.6258 .% -0.6258 1.521%Feedwater Temperature

,15,62682

-6.9610 -,,. _ 6.9060 Feedwater.

Pressure 0.0884 .-0.000941 -a.......

.0002 0.0 Steam Pressure.

r40Q837 -0.4980 -1.1812 1.67912 4.082%Plant Coffipiiter Complthtion of Enthalpies IJ 60370 "-0:031 0.0.21 0.07 8%.Other Inpuis __________

_________

-6:4657 -6.4657 i5.7189/%Totals __________

-33.4895 1 7.6469 -41.17365 i00.000%:CCN-IC-08001 Rev.00 P~igi 28ýof CA06945; Revisibhn0000

ýCALORIMETRIC UNCERTAINTY USING TIlE L 2FMCHEC-KPLUS FLOW 'MEASUREMENT SYSTEM

8.0 CONCLUSION

This calculation9determines~th1 calorimetric uncertainty using the Caldon LEFM'8CieckPlis uiltiaso'nic.flowý nIeasurcmn-ent system, to measure fedxvater flow. Un'Ceitinity'is`evAlu ated af thc proposed.Appe'ndix K u0pr'ted pow&rt, of2737 MW(th)*. 'The Appendix K power representsaii increase 6fapproximately 1.4%from pthe current licensed pw limit of2a700MW(oth):.

This calculation contains eva'ri-oii's'ifveiierified as designprogresseýs for ihe. UFs: Specifically,assu'rfipti61ons 5.1ý2,.5:2];

and:5.2.3 will be Verfie&lasa ppart ofthe4design.process:

Additi6nally, the oriservatism tof the a6sunmptiorns iegardinrg maximum and'.minimumrhuprattd 1'aint proc6e6s parameters'to be used for the computationh (5,4.1 and 5.4.2) will be verified upon statup aftetbel'oer uprate. H ,ey.,i, tlic piocess forwhichthese wer" wcosen was vyey co!se',atiyv, and tl:!e'iiilyzed limits are not anticipated'tochange.

Results Uýihg: Sinel eý-Side-of-Inteest:AAproaich::

'This calculation determines total secondary calorimetric calculation uncertainties using standard.methodology. calculafion is performed in general accordance with Reftereice 6.1, c6nsidering.

the single-side-6f-interet approach, as defined in Section'81 offRef{reince 6.11. The.following tabie*presetnts, the results of this analysis, where is the total secondary calorimetric uncertainty.

ARated]ppendixK Thermal UAL.NI(SS)

Margin 'Margin Liit Power MW MW %RTP 2754 '.2737 -1240559 .9441 0:18066%Since.adeqUateiai-gin e~ists-between the propos'ed R'atd Thermal P6owierand-ihe AppendikiK limfitt6 account fur iiiastuhihnt'huicerthinty uidadditiui al' i iagirn.tthe propued TheriiiPue'l ii deemnied acCeptiableý, cofnsidering ,in.strument rincertainity:

The relative conitributio'nsofeaich ncertainty termf to thetýotal se'on'dar'ycalorimeiiic uncertainty are provided in&tlie table below.UINPUT MBTU/hr (U /h,.r( s)) P Ur/MBTU/hr JQT/h INPUT Feedwater Flow. -29. [498 -25:3725 Blowdown Flow -4.5779 -0.6258 Feedwater "Femperature

-15:2682 --6.9610 reedwaterPressure -b.60 4 -o0`6002 Steam., Press re -4:0837 -0.4980 Plant:C6hipiiui Cdomputation of Entlialpies

.1.0360 -0.0321 (U, NPUT-INET(SS))

UN1TNTS) MBUhr IPTNTS)s~ --25.3725 61,6799%-0.6258 1.521%-6.96 10 16.9220%-0.0002 0.001%1-1.6792 4.082%20.. .032, 0.078%S -6.4657 151718106-41 .13065 100.000 Other Inbuts Totals 1:, -33.4895 -7 CCN-ICG08001 RPev. 0a Page 29 6Of 30 CA06945,Revision 00,00... .CA~i:R{IMETRid UNcERTA1NTUUNG THE LE0M Ci IECKPLUS t:.w M EASRElENTrSYSTEM Results if SiI igle-Side-of-Interest Approach is Not Credited: Thiis calculation-also.

presents,the sarne'setof results as above for the.casetwhere the'single-side-of-inrterest:

approach, as-defined'in Section 8.1 of Reference 6.1 1i]s not credited.

Uncder this case, the foilowing table: presents the results of this analysis; where UCALNFT is:the total secondary'calorimetric uncertainty.

Append ixlK Rated Appendix UC t. A;.NF Margin .Margin Limit ,Toerma mW M.W %RTP Power.2754 ý2737 -1319353 ,3.01647 ,0.'1120%Since adequatemargin exists between the~proposed Rated THermal Powerandthe Appe ndixK'limit't6.

account for- instrument uncertainty and additional margin, the~proposed Rated Thermal Power limitvis deemed acceptable,, considering instrument uncertainty'.

The relative contributions of eachuncertaInty term tothe total secondar9 cal 6#imetricuncertainty.

are provided in the:table below........ (U ITN, NUUT.NETNT INPUT MBTU/hr MBTU/hr MBTlU/hr, MBT-,-U/brN Feedwater Flow -34.7317 -30.2310 , -30'.2310 631578%Blowdown Flow -5A545 -0:7456 -6 7-.7456% 1.568%Fcedwatcr Tcmocrathrc

-18.1920 -8:2!)39 " ,82939 17.443%Feedwater Pressure, -0.1053 -0.0003 -0:00030.001%

" Steam Pressure -4.865,7 -0.5-33 1,1812 -1.7746 3:732%Planl Comiputer CompuItatioin

.-03. __oIfEnthalpis.

-1 .2356 -0.0383 _o __ _ -0 .03,83_ 0.080%O ther Input ,, .., : -6 4657 ,,, -6.4657 13 -598%Totals -39.9024 1 7'6469 -47.5493, 1I00.000A.

CCN-IC-0800J1 Rcv. g Page! 3 01.of 3.0 ATTACHMENT (3)MARKED UP TECHNICAL SPECIFICATION PAGES Renewed Operating License Page 3 (Unit 1)Renewed Operating License Page 3 (Unit 2)1.1-5 Calvert Cliffs Nuclear Power, Inc.August 29, 2008 rules, regulations, and orders of the Commission, now or hereafter applicable; and is subject to the additional conditions specified and incorporated below: (1) Maximum Power Level -73 The licensee is authorized to operate th facility at steady-state reactor core power levels not in excess of 2-1-e&megawatts-thermal in accordance with the conditions specified herein.(2) Technical'Spdcifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 286, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications.(a) For Surveillance Requirements (SRs) that are new, in Amendment 227 to!Facility Operating License No. DPR-53, the first performance is due at the end of the first surveillance interval that begins' at implementation of Amendment 227. For SRs that existed prior to Amendment 227, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is-due at the, end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 227.(3) Additional Conditions The Additiona~l Conditions contained in Appendix C as revised through Amendment No. 267 are hereby incorporated into this license. Calvert Cliffs NucleariPower Plant, Inc. shall' operate the facility in accordance with the Additional Conditions.

(4) Secondary Water Chemistry Monitorinq Progqram The Calvert Cliffs Nuclear Power Plant, Inc., shall implement a secondary, water chemistry monitoring program to inhibit steam generator tube degradation.

This program shall include: a. Identification of a sampling schedule for the critical parameters and control points for these parameters;

b. Identification ofthe procedures used to quantify parameters that are critical to control points;Amendment No. 286 C. This license is deemed to contain and is subject to the conditions set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act, and the rules, regulations, and orders of the Commission, now andhereafter applicable; and is subject to the additional conditions specified and incorporated below: (1) Maximum Power Level -737 The licensee is authorized to operate th facility at reactor steady-state core power levels not in excess of 2-7--e&megawatts-thermal in accordance with the conditions specified herein.(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 263 are hereby incorporated into this license.The licensee shall operate the facility in accordnace with the Technical Specifications.(a) For Surveillance Requirements (SRs) that are new, in Amendment 201 to, Facility Operating License No. DPR-69, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 201. For SRs that existed prior to Amendment 201, including SRs with modified acceptance criteria and SRs whose frequency of performance is being. extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 201.(3) Less Than Four Pump Operation The licensee shall not operate the reactor at power levels in excess of five (5) percent of rated thermal power with less than four (4) reactor coolant pumps in ope'ration.

This condition shall remain in effect until the licensee has submitted safety analyses for less than four pump operation, and approval for s:uch operation has been granted by the Commission by amendment of this license.(4) Environmental Monitoring Pro gram If harmful effects or evidence of irreversible damage are detected by the biological mo nitoring program, hydrological monitoring program, and the radiological monitoring program specified in the Appendix B Technical Specifications, the licensee will provide to the staff a detailed analysis of the problem and a program of remedial action to be taken to eliminate or significantly reduce the -detrimental effects or damage.Amendment No. 263 Definitions 1.1 1.1 Definitions OPERABLE-OPERABILITY PHYSICS TESTS A system, subsystem, train, component,.or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s), and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

/PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are: a. Described in Chapter 13, Initial Testsland Operation of the Updated Final Safety Analysis Report;b. Authorized under the provisions of 10CFR 50.59; or!c. Otherwise approved by the Nuclear Regulatory Commission.

RTP shall be a total reactor core heat transfer rate to the reactor coolant of.-2-*& MWt.RATED THERMAL POWER (RTP)REACTOR PROTECTIVE SYSTEM (RPS) RESPONSE TIME CALVERT CLIFFS -UNIT 1 CALVERT CLIFFS -UNIT 2 The RPS RESPONSE TIME shall be that time interval from when the monitored parameter.

exceeds its RPS trip setpoint at the channel sensor until electrical power to the CEAs drive mechanism is interrupted.

The response time may be measured b!means of any series of sequential, overlapping, oi total steps so that the entire response time 'is measured.

In lieu of measurement, response time may be verified for selected components provided that the components and methodology for r 1.1-5 Amendment No. 286 Amendment No. 263.